Gas IN11 1 Electric Company - NRC: Home Page › docs › ML0312 › ML031260681.pdfFinal Draft All discrepancies resolved with Lead Licensing Engineer * Enter N/A where not applicable

IN11 1Pacific Gas andElectric Company

David H. OatleyVice President andGeneral Manager

Diablo Canyon Power PlantPO Box 56Avila Beach, CA 93424

805 545 4350Fax 805 545 4234

April 30, 2003

PG&E Letter DCL-03-049

U.S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001

Docket No. 50-275, OL-DPR-80Docket No. 50-323, OL-DPR-82Diablo Canyon Units 1 and 22002 Annual Radioactive Effluent Release Report

Dear Commissioners and Staff:

PG&E is submitting the enclosed 2002 Annual Radioactive Effluent Release Reportin accordance with 10 CFR 50.36a (a)(2) and Section 5.6:3 of the Diablo CanyonPower Plant Technical Specifications.

The report describes the quantities of radioactive gaseous and liquid effluentsreleased from the plant, and the solid radioactive waste shipments made during theperiod of January 1 through December 31, 2002.

One compact disk is being sent with the report. The disk contains meteorologicaldata. If you have any questions, please contact Jeff Gardner of my staff at(805) 545-4385.

Sincerely,

David H. Oatley

swh/3649/R0232137Enclosurecc/enc: Edgar D. Bailey, DHS

RogerW. BriggsHarvey CollinsEllis W. MerschoffWilliam A. NestelDavid L. ProuIxGirija S. ShuklaGregory ThomasDiablo Distribution

A member of the STARS (Strategic Teaming and Resource Shanng) Alliance

Caltaway . Comanche Peak * DiabLo Canyon . Palo Verde . South Texas Project * Wolf Creek

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.6 - SUBMITTAL PROCESSING CHECKLISTor iA

PG&EiLetter No. (e.g., DCL, DIL, etc. DCL.03-049

Subject: 2002 Annual Radioactive Effluent Release Report

Target Submittal Date 28 APR 2003 Firm Submittal Date: 30 APR 2003 N/A C]

File Location: S:\RS\ RA\2002 Annual RERR Report Rev 1.docInitial / Date

1. References & bases identified for factual information.

2. Outgoing Correspondence Screen (RS-2) completed & commitments captured per

Xl4.ID1.

3. RS Manager's concurrence for release obtained.

4. Record of Review Checklist completed & signed per XI1.ID1.

5. Clerical reviews completed. Draft # I

Draft# - Final (on signatory letterhead)

6. Peer Review Checklist (RS-4) completed

7. Provide this checklist, the final letter & enclosures, Record of Review Checklist, &

Outgoing Correspondence Screen to signatory.. For FIRM submittals, was 2 days met? Yes E No D N/A I

8. WNill entire submittal, including enclosures, be placed in RS & ACTS libraries and

EDMS? I Yes Cl No

If NO, complete form RS-IA, indicating where the enclosures, etc. will be maintained.

9. Indicate RMS access level. Check one box only.

[ X] Reg Regular[ "D" Proprietary

["B" Personnel Record[ ] "A" Security Safeguards Information (SSI)

* RMS folder name(i.e., LAR, LER, etc.)

* Put in envelope with a copy of this form attached on outside and put inSSI safe in RS Managers office.

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10. Place copy of this checklist (including RS-lA, if applicable), the submittal & enclosures,

Record of Review Checklist, & the Outgoing Correspondence Screen in the RMS

I (fireproof) file cabinet (or the SSI safe, if applicable).

11. Deliver this checklist (including RS-1A, if applicable), original of signed submittal &

enclosures, Outgoing Correspondence Screen, & Record of Review Checklist to clerks.

12. Tracking documents (i.e., NCR ACTs or AEs) have been updated for submittalcompletion (including the LERtemplate.ppt, if the submittal is an LER). N/A [9

* LERtemplate.ppt is located at S:\RS\DCISC\LERs\LERtemplateYYYY.ppt

13. Verify commitment entry on NCR ACTs or AEs within 15 days. N/A ER

so)J-k I -4 0

~5 .t-jri / 4--

RS-1

/-

ISUBMITTAL PROCESSING CHECKLIST - Supplement

Identify the enclosures, etc. (e.g., proprietary, personal, or SSI) that will NOT be filed in the RS

and ACTS libraries, or EDMS, and where the actual documents will be maintained.

Examples:Enclosure 1 NRC 396 forms are retained by Learning Services and the DCPP

(Personal information) Medical Facility.Enclosure 3 The diskettes are maintained by Radiation Protection.

(Diskette)Entire Document The complete proprietary version of this document is available in the

PG&E Law Department in San Francisco.

Item LocationEnclosure: TES Data maintained by TES in San RamonMeteorological Data:CD-R

Attachment 1: Maintained in EDMSPD CY2, Rev 5

Attachment 2: Maintained in EDMSIDAP CY2.lD1,Rev. 6A

Attachment 3: Maintained in EDMSCAP A-8, Rev. 26

Attachment 4: Maintained in EDMSDLAP RP2.DC2,Rev. 12, U

Attachment 5: Maintained in EDMSDLAP RP2.DC3,Rev.Ws'

RS-1A

I-

. -/

I 07/23/02NUCLEAR POWER GENERATION

XI1.ID1ATTACHMENT 8.1

Page 1 of 1

TITLE: Regulatory Submittal Record of Review Checklist

SUBMITTAL TITLE Zeros ~41 #CoSitLchct d 407-

REQUIRED SUBMITTAL DATE:

PRIMARY REVIEWERS

LEAD TECHNICAL REVIEWER

LEAD TECHNICAL DIRECTOR/1

LICENSING MANAGEMENT

INDEPENDENT TECH REVIEWE

CROSS-DISCIPLINE REVIEWER(

[ ] OPERATIONS

[ ] MAINTENANCE

[ ] ENGINEERING

[ ] PLANT SUPPORT

[ ] QUALITY

[ ] LAW

[ ] TES

[ ]CHEMISTRY & ENVIROI

a> neA I d [] N/A SCHEDULED SUBMITTAL DATE:

NAME COMMENTSYES NO

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_ _ _ _ _ _ _ _ _ _ _ _ [] []

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__ _ _ _ _ _ _ _ _ _ _ [] []

_ _ _ _ _ _ _ _ _ _ _ _ [] []

[__ _ _ _ _ _ _ _ _ _ ] [ 1[__ _ _ _ __ _ _ _ _ ] []

>4MENTAL [] []

__ __ __ __ __ __ __ _ [] []

______________[ ] []

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RESOLVED

YES NO[]i [1

[] [I][] [][] [][] [][] [][] [][] [][] [][] [1] I[]

[II [III] II[I [I

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OPERATIONSPSRCNSOCOTHER

COMMENTS RESOLVEDSECONDARY REVIEWERS NAME COMMENTS

LAWC-Z-/ e ALLA/ -f-1-

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YES NO

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CONCURRENCE HAS BEEN RECEIVED FROM PRIMARY REVIEWERS AND TECHNICAL COMMENTSTT A 170 flrVXT DTVQT VVT1oIEAD LICENSIN ENI N, - R

LEAD LICENSING ENGNEER: ER-. DATE: 4_ _ __ ___

01317407.DOC 01A 0813 0656

I *

- PEER REVIEW CHECKLISTLICENSING SUBMITTAL

i (To be performed on final draft only.)

|PG&E Letter No. (e.g., DCL, DIL, etc.) -IJI I-

ITEM DESCRIPTION INITIALS|Cover Letter Correct signatory letterhead (Ref. MUD)

Full names used for signatory & cc list; right people listedTitle correctLetter number verified against outgoing correspondence logLetter number appears on all pagesAll pages numbered, except first pageDate correct and appears on all pages (month, day, and year)Address and docket number(s) correctText reviewed for obvious errorsTS and/or 10 CFR references correctIf affidavit required or NOV response, verify Law Department has reviewed

All Submittals Text reviewed for obvious errorsRevision bars included (if applicable)TS and/or 10 CFR references correctReferences to other documents correct (e g, DCLs, FSAR, etc.)Submittal addresses the specific regulation requirementsEnclosures labeled

Outgoing Letter number and title correctCorrespondence FSAR Update Review- one box checked.Screen (RS-2) Commitment implemented before or after LA receipt (LARs/RAls only) at

Commitment(s) quoted verbatim (& clarifications made if needed) W fTracking Document - AR/AE or NCR Action numbers -.Assigned To - Name & Organization CodeCommitment Type - Firm or Target & Due Date AOutage Commitment - Y or N indicator & Applicable OutagePCD Commitment - Y or N indicator & Implementing DocumentsDuplicate of NCR Commitment in NCR - Y or N indicator & PCD numberAn individual from all departments assigned commitments was a Cross-Discipline Reviewer (Ref. Xl1.ID1, Step 5.2.3.a 4.a)

LER Forms LER number correct; consistent with cover letterLER number & docket number(s) on first & remaining pagesTitle consistent with cover letter & Outgoing Correspondence Screen (RS-2)Dates correct on first page header (month, day, & year)Dates & times consistent with 10 CFR 50 72 ENS reports madeDates & times consistent with other source documentsDates & times consistent between abstract and narrativePage numbers correct & all pages accounted forAbstract word count <1400 characters (including spaces)IEEE 803 codes entered and correct

Procedure Submittals Procedure revision numbers current using EDMS (e.g ,'EPIP)Filing Instructions Filing instructions clear (Per RS-1A, if applicable)

ROR Checklist Record of Review Checklist completed and signedFinal Draft All discrepancies resolved with Lead Licensing Engineer

* Enter N/A where not applicable.

I have reviet 2 issi tl\for thee item*Ritialed above. This submittal is ready f the/natory.

!E, n yl§~//2 En V,, v, v -_ . -

TPe'~fromrrfi~el~y ,/ DatW

RS-4

bcc/enc*: David C. ChenJeffery E. GardnerJohn R. KnerneyerRobert W.- Lorenz

* To save some paper, bcc's should receive the enclosure withoutattachments 1-5 (copies of DCPP procedures).

OUTGOING CORRESPONDENCE SCREEN(Remove prior to NRC submittal)

Document: -PG&E Letter DCL-03-049Subject: 2002 Annual Radioactive Effluent Release ReportFile Location S:\CHEM\SECURE\Annual Rad Eff Report\2002\2002 Annual

Report Draft Rev 1.DOCVersion:

FSAR Update ReviewUtilizing the guidance in X13.1D2, does the FSAR Update need to be revised? Yes LI No 1lIf 'Yes, submit an FSAR Update Change Request in accordance with X13.1D2 (or if this is an LAR, process in accordance with WG-9J

Commitment -#1-Statement of Commitment: NONE

DIABLO CANYON POWER PLANT2002 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT

January 1 - December 31, 2002

EnclosurePG&E Letter DCL-03-049

DIABLO CANYON POWER PLANT

Annual Radioactive Effluent Release ReportJanuary 1. 2002 Through December 31, 2002

Table of Contents

Introduction ............................................... 3

I. Supplemental Information .4

II. Major Changes to Liquid, Gaseous, and Solid Radwaste Systems .10

Ill. Changes to Radiological Monitoring and Controls Program (RMCP) .10

IV. Changes to the Radioactive Effluent Controls Program (RECP) .10

V. Changes to the Off-Site Dose Calculation Procedure (ODCP) .10

VI. Changes to the Process Control Program (PCP) .1

VII. Land Use Census .11

Vil. Gaseous and Liquid Effluents .............................................. 12

TABLE I - GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES .............................................. 12TABLE 2 - GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES .................................................... 15TABLE 3- GASEOUS EFFLUENTS - LOWER LIMITS OF DETECTION ................................................ 19

TABLE 4- LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES ................................................... 21TABLE 5- LIQUID EFFLUENTS - NUCLIDES RELEASED ............................................................. 24TABLE 6- LIQUID EFFLUENTS - LOWER LIMIT OF DETECTION (LLD) .............................................. 30

IX. Solid Radwaste Shipments .33

X. Radiation Dose due to Gaseous and Liquid Effluents .36

TABLE 7- RADIATION DOSE DUE TO THE RELEASE OF RADIOACTIVE LIQUID EFFLUENTS 38TABLE 8- RADIATION DOSE DUE TO THE RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS .39

TABLE 9- PERCENT OF TECHNICAL SPECIFICATION LIMITS FOR RADIOACTIVE LIQUID EFFLUENTS .42

TABLE 10- PERCENT OF TECHNICAL SPECIFICATION LIMITS FOR RADIOACTIVE GASEOUS

EFFLUENTS .43

TABLE 11 - ON-SITE DOSE TO MEMBERS OF THE PUBLIC .45

XI. Meteorological Data .50

Page 1



Annual Radioactive Effluent Release ReportJanuary 1, 2002 Through December 31, 2002

Attachments

1. Radiological Monitoring and Controls Program (RMCP), CY2 Revision 5

2. Radioactive Effluent Controls Program (RECP), CY2.0D1 Revision 6A

3. Off-Site Dose Calculation Procedure (ODCP), CAP A-8 Revision 26

4. Radwaste Solidification Process Control Program (PCP), RP2.DC2 Revision 11(incorporates revision 8-11)

5. Radwaste Dewatering Control Program, RP2.DC3 Revision 5

6. 2002 Land Use Census

Page 2



2002 Annual Radioactive Effluent Release Report

Introduction

The 2002 Annual Radioactive Effluent Release Report summarizes gaseous and liquideffluent releases from Diablo Canyon Power Plant's (DCPP) Units 1 and 2. The reportincludes the dose due to release of radioactive liquid and gaseous effluents andsummarizes solid radwaste shipments. The report contains information required by Units 1and 2 Technical Specification (TS) 5.6.3 and is presented in the general format ofRegulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in SolidWastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Nuclear Power Plants," Appendix B, "Effluent and Waste Disposal Report."Procedure revisions, which implement the Radiological Monitoring and Controls Program(RMCP), Radioactive Effluent Controls Procedure (RECP), Off-Site Dose CalculationProcedure (ODCP), and Process Control Program (PCP), and one compact disk containingmeteorological data, are attached.

In all cases, the plant effluent releases were well below TS limits for the report period.

Page 3


Supplemental Information

A. Regulatory Limits

1. Gaseous Effluents

a. Noble Gas Dose Rate Limit

The dose rate in areas at or beyond the site boundary due toradioactive noble gases released in gaseous effluents is limitedto less than or equal to 500 millirem (mR) per year to the totalbody and less than or equal to 3000 mR per year to the skin.(Radioactive Effluent Controls Program 6.1.6.1.a.)

b. Particulate and Iodine Dose Rate Limit

The dose rate in areas at or beyond the site boundary due toiodine-131, iodine-133, tritium, and all radionuclides inparticulate form with half lives greater than 8 days in gaseouseffluents, is limited to less than or equal to 1500 mR per year toany organ. (RECP 6.1.6.1.b)

c. Noble Gas Dose Limit

The air dose due to noble gases released in gaseous effluentsfrom each reactor unit to areas at or beyond the site boundary islimited to the following:

Radiation Type Calendar Quarter Limit Calendar Year LimitRECP 6.1.7.1.a RECP 6.1.7.1.b

Gamma 5 millirad 10 milliradBeta 10 millirad 20 millirad

d. Particulate and Iodine Dose Limit

The dose to an individual from iodine-1 31, iodine-1 33, tritium,and all radionuclides in particulate form with half lives greaterthan 8 days in gaseous effluents released from each reactor unitto areas at or beyond the site boundary is limited to less than orequal to 7.5 mR to any organ in any calendar quarter and lessthan or equal to 15 mR to any organ during a calendar year.(RECP 6.1.8.1)

2. Liquid Effluents

a. Concentration

The concentration of radioactive material released from the siteis limited to the concentrations specified in 10 CFR Part 20,Appendix B, Table 2, Column 2, for radionuclides other thandissolved or entrained noble gases. For dissolved or entrainednoble gases, the concentration is limited to 2 x 104microcuries/milliliter ([iCi/ml) total activity. (RECP 6.1.3.1)

Page 4


b. Dose

The dose or dose commitment to a member of the public fromradioactive materials in liquid effluents released from eachreactor unit to areas at or beyond the site boundary is limited tothe following:

Dose Type Calendar Quarter Limit Calendar Year LimitRECP 6.1.4.1.a RECP 6.1.4.1 .b

Total Body 1.5 millirem 3 milliremAny Organ 5 millirem 10 millirem

B. Maximum Permissible Concentrations


Maximum permissible concentrations are not used for determiningallowable release rates for gaseous effluents at DCPP.

2. Liquid Effluents

The concentrations listed in 10 CFR 20, Appendix B, Table 2, Column2, for radionuclides other than dissolved or entrained noble gases areused for determining the allowable release concentration at the point ofdischarge from the site for liquid effluents. For dissolved or entrainednoble gases, the allowable release concentration at the point ofdischarge is limited to2 x 10-4 gCi/ml total activity for liquid effluents.

C. Measurements and Approximations of Total Radioactivity

1 . Gaseous Effluents

a. Fission and Activation Gases

A pair of off-line monitors equipped with beta scintillatordetectors monitors the gaseous radioactivity released from theplant vent. The monitor readings are correlated to isotopicconcentration based on laboratory isotopic analysis of grabsamples using a germanium detector.

For plant vent noble gas releases, grab sample results are usedto quantify releases. The individual batch release data are usedto quantify the radioactivity discharged from the gas decay tanksand containment.

A noble gas grab sample is obtained and analyzed at leastweekly. The isotopic mixture is assumed to remain constantbetween grab sample analyses.

Page 5


Containment purges, gas decay tank releases, and air ejectordischarges are released via the plant vent.

The gaseous radioactivity released from the steam generatorblowdown tank vent is measured by analyzing liquid or steamcondensate grab samples with a germanium detector. A factorR, a ratio of unit masses between water flashing to steam andwater entering the tank, is used to calculate the activity. Theisotopic concentrations are assumed to remain constantbetween grab samples.

Other potential pathways for releasing gaseous radioactivity areperiodically monitored by collecting grab samples and analyzingthese samples with a germanium detector system.

b. lodines

Radioiodines released from the plant vent are monitored bycontinuous sample collection on silver zeolite cartridges. Thecartridges are changed at least weekly and analyzed with agermanium detector. The radioiodine releases are averagedover the period of cartridge sample collection.

Other potential pathways for releasing radioiodines areperiodically monitored by collecting samples using charcoal orsilver zeolite cartridges and analyzing with a germaniumdetector.

Radioactive materials in particulate form released from the plantvent are monitored by continuous sample collection onparticulate filters. The filters are changed at least weekly andanalyzed with a gernianium detector. The particulateradioactivity is averaged over the period of particulate filtersample collection. Each filter is analyzed for alpha emittersusing an internal proportional counter. Plant vent particulatefilters collected during a quarter are used for the compositeanalysis for strontium-89 and -90, which is counted on aninternal proportional counter after chemical separation.

Other potential pathways for releasing radioactive particulate areperiodically monitored by collecting samples using particulatefilters and analyzing these filters with a germanium detector.

c. Tritium

Tritium released from the plant vent is monitored by passing ameasured volume of plant vent sample through a water columnand determining the tritium increase in the water. An aliquot ofthe water is counted in a liquid scintillation spectrometer. Theminimum routine sample frequency for tritium is weekly. Thetritium concentration is assumed to remain constant betweensamples.

Page 6


d. Estimations of Overall Error

Sources of error considered for batch release are: (1) calibrationsource; (2) calibration counting; (3) sampling; (4) samplecounting; and (5) gas decay tank pressure gauge / containmentexhaust fan flow rate.

Sources of error for continuous release are: (1) calibration source;(2) calibration counting; (3) sampling; (4) sample counting; (5)process monitor (RE-14) reading (fission gases only); and (6)plant vent exhaust fan flow rate.

Total error = (a 21 + G2 2 + a23+ c 2;) 112

Where a, = error associated with each component

.2. Liquid Effluents

a. Batch Releases

Each tank of liquid radwaste is analyzed for principal gammaemitters using a germanium detector prior to release. A monthlyprerelease analysis includes dissolved and entrained gases.Volume proportional monthly and quarterly composites areprepared from aliquots of each tank volume discharged. Themonthly composite is analyzed for tritium using a liquidscintillation spectrometer and gross alpha radioactivity using aninternal proportional counter. The quarterly composite isanalyzed for iron-55 using a liquid scintillation spectrometer andfor strontium-89 and -90 using an internal proportional detectorfollowing chemical separations.

b. Continuous releases

For the continuous liquid releases of the steam generatorblowdown tank and turbine building sump oily water separator,daily grab samples are collected and aliquots are proportionedfor weekly, monthly, and quarterly composites.

The oily water separator weekly composite is analyzed forprincipal gamma emitters using a germanium detector. Thesteam generator blowdown tank weekly composite is analyzedfor principal gamma emitters and iodine-131.

The steam generator blowdown tank monthly composite isanalyzed for tritium using a liquid scintillation spectrometer andfor gross alpha using an internal proportional counter.

The steam generator blowdown tank quarterly composite isanalyzed for iron-55 using a liquid scintillation spectrometer andfor strontium-89 and strontium-90 using an internal proportionalcounter following chemical separation. The results for each ofthe composites are averaged over the period of the composite.

Page 7


In addition, one grab sample of the steam generator blowdowntank is analyzed monthly for dissolved and entrained gasesusing a germanium detector. The results of this analysis areassumed to remain constant over the period of one month.

A grab sample of the steam generator blowdown is collected atleast weekly and analyzed for gamma activity using agermanium detector. This analysis is used to monitor activity,however, is not used in effluent calculations unless a significantchange is detected.

Note on dilution volume:

Tables 4A, 4B and 4C, "Liquid Effluents - Summation Of AllReleases," item F., lists the "Volume of circulating saltwater usedduring release periods" in liters. This value is calculated bymultiplying the discharge duration by the circulating water flowrate. The values listed in the Tables are the summation of thecirculating water discharge volume calculated for each individualbatch and continuous discharge period. Therefore, in the casewhere two or more simultaneous discharges into the samecirculating water are occurring, the calculated volume ofcirculating water is duplicated, and therefore the sum of thedilution volumes for the batch releases and continuous releasesare greater than the actual dilution volume since each dischargeincorporates the circulating discharge flow rate in its own dosecalculation.

c. Estimation of Overall Error

Sources of error considered are: (1) calibration source error; (2)calibration counting error; (3) sampling error; (4) samplecounting error; and (5) volume-of waste release error.

These sources of error are independent; therefore the total erroris calculated according to the following formula:

Total error = (02l + C22 + U2 3 + -- 2 2) 1/2

Where a = error associated with each component

Page 8


D. Batch

1.

Releases

Liquid

a. Number of batch releases......................................

b. Total time period for batch releases........................

c. Maximum time period for a batch release.................

d. Average time period for a batch release.....................

e. Minimum time period for a batch release....................

f. Average saltwater flow during batch releases.............

613

2870 hours

168 hours

4.68 hours

0.02 hours

9.27E+05GPM

2. Gaseous

a. Number of batch releases.......................................

b. Total time period for batch releases...........................

c. Maximum time period for a batch release..................

d. Average time period for a batch release..................

e. Minimum time period for a batch release.................

130

1608 hours

72 hours

12.37 hours

1.00 hours

E. Abnormal Release (Gaseous and Liquid)

1. Leakage of waste gas from the Waste Gas Header occurred between10/11/02 - 10/13/02. The gas leaked through the boric acid evaporatorsystem into the Auxiliary Building, and eventually through the PlantVent. The leak in the boric acid evaporator system was isolated to stopthe un-planned, monitored discharge.

The release rate of noble gas to the plant vent resulted in 0.58% of therelease rate limit. The resulting gamma air dose and beta air dose fromthis discharge was 1.42E-3 mrad and 4.99E-4 mrad, respectively.

2. An unplanned, monitored discharge was made from the WasteHandling And Treatment system. A valve alignment was made thatallowed water from one of the tanks to inadvertently discharge.

Page 9


The radioactive concentration of the discharged water was 6.01 E-5pCi/ml. The effluent concentration limit for this discharge was 2.55E-4pCi/ml. Therefore, the discharge was below the discharge limit.

The total body dose for this discharge was 1.84E-8 mrem. The highestorgan dose was 3.17E-6 mrem (thyroid).

3. The Liquid Radwaste (LRW) monthly composite sample for June 2002was analyzed for gross alpha. The analysis result was <1.1 9E-7pCi/ml. The required lower limit of detector for this analysis is 1.OE-7pCi/ml. Therefore, the value included in this report is listed as 1.19E-7pCi/ml. This is reflected on Tables 4A and 4C, Section D.

4. A summation error was identified on Table 4D, Section D. The "AnnualTotal" curie value for Liquid Effluent Gross Alpha only reflected theSecond Quarter value from Table 4A. The summation for this report(2002 data) is correct. Any corrections to previous year's reports due tothis error will be included in next year's Annual report.

II. Major Changes to Liquid, Gaseous and Solid Radwaste Treatment System

There were no major changes to liquid, gaseous, and solid radwaste treatmentsystems during the report period.

Ill. -Changes to the RMCP

CY2, "Radiological Monitoring And Controls Program," was revised during the reportperiod.

Revision 5 made the following changes:

* Changed the frequency of the Radioactive Effluents Controls Program auditsfrom 12 months to 24 months to reflect a recent change in the FSAR.

* Updated the procedure to convert position titles to a generic format.

Revision 5 was reviewed by the Station Director on September 25, 2002, andapproved by the Chief Nuclear Officer on October 2, 2002. (See Attachment 1)

IV. Changes to the RECP

CY2.lD1, "Radioactive Effluents Controls Program," was revised during the reportperiod.

Revision 6A was a minor editorial correction that did not change the technical contentof the procedure. This change was effective on February 1, 2002. The change didnot require review or approval from the Station Director. (see Attachment 2)

V. Changes to the ODCP

CAP A-8, 'Off-Site Dose Calculations Procedure," was revised. Revision 26 madethe following change:

Updated the gaseous effluent dispersion and deposition values (X/Q and D/Q)based upon the most recent 5 year average annual meteorological conditions.

Page 10


Revision 26 was reviewed by the PSRC and approved by the Station Director onMay 31, 2002. (See Attachment 3)

VI. Changes to the Process Control Program

RP2.DC2, "Process Control Program," was revised five times during the reportperiod.

A. Revision 8 made the following change:

* The title cited in the procedure for this annual report.

* The Station Director approved revision 8 on March 25, 2002.

B. Revision 9 made the following change:

Advance polymer solidification agent was added to the process controlprogram.

* The Station Director approved revision 9 on July 3, 2002.

C. Revision 10 made the following change: I

* The advanced polymer was altered from a 4-component to a 3-component mixture to achieve an extended gelation time.

* The Station Director approved revision 10 on September 30, 2002.

D. , Revision 11 made the following change:

* An altered 4-component polymer was adopted for extended gelationbased upon a specimen result of the 3-component mix.

* The Station Director approved revision 11 on October 17, 2002. (SeeAttachment 4)

RP2.DC3, "Dewatering Control Procedure" was revised once during the reportperiod

A. Revision 5 made the following change:

* Allows Class A waste above lpCi/cc to be loaded into steel liners,versus HICs. This matches the new trench at the Envirocare site.

The Station Director approved revision 5 on February 8, 2002. (SeeAttachment 5)

VII. Land Use Census

Changes to the Land Use Census Program are included as Attachment 6.

Vil. Gaseous and Liquid Effluents

Tables 1 through 3 describe gaseous effluents. Tables 4 through 6 describe liquideffluents.

Page 11



ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002TABLE 1A

GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES

i First Second Est.TotalI Unis |Quarter __Quarter Error %

A. Fission and activation gases

1. Total release Ci 1.08E+1 1.52E+O 46% |

2. Average release rate for period piCi/sec 1.39E+O 1.93E-1

3. Percent of technical specification limit' 3.57E-3 I 1.40E-3

B. lodines

1. Total iodine-131 Ci 1.71 E-4 1.76E-5 24%

2. Average release rate for period ItCi/sec 2.20E-5 2.24E-6

? 3. Percent of technical specification limit' 1.31 E-4 1.33E-5

C. Particulates

1. Particulates with half-lives >8 days Ci 5.04E-4 2.65E-4 24%

2. Average release rate for period pCi/sec 6.48E-5 3.37E-5

3. Percent of technical specification limit' 2.35E-5 1.37E-5

4. Gross alpha radioactivity Ci 4.46E-7 1.35E-7

D. Tritium

1. Total release Ci | 3.09E+1 [ 1.13E+2 0

________________________________________________ ___________ __________I

2. Average release rate for period pCi/sec 3.97E+O 1.44E+1

3. Percent of technical specification limit' I % I 9.33E-6 I 3.38E-5 |

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unitmass or volume). This note applies to all tables.

' RECP 6.1.6.1 Limit

Page 12



ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002

TABLE IB

GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES

I Third | Fourth Est.Totail| U Quarter Quarter | Error %


1. Total release Ci 7.58E-1 3.72E+O 46%

2. Average release rate for period pCI/sec 9.54E-2 4.68E-1

3. Percent of technical specification limit: 1.IIE-4 1.20E%3

B. lodines

1. Total iodine-131 Ci 5.50E-5 2.12E-3 24%l

x 2. Average release rate for period piCilsec 6.92E-6 2.67E-4

1 3. Percent of technical specification limit' % 1.46E-5 1.51 E-3

C. Particulates

1 Particulates with half-lives >8 days Ci 2.35E-5 1.25E-4 24%

2. Average release rate for period pCi/sec 2.96E-6 1.57E-5

3. Percent of technical specification limit' - 1.20E-6 5.24E-6

4. Gross alpha radioactivity Ci MDA MDA

D. Tritium

1. Total release Ci 5.18E+1 1.11E+2 13 |

2. Average release rate for period IiCi/sec 6.51 E+0 1.40E+1

3. Percent of technical specification limit' I % I 1.53E-5 I 3.28E-5


'RECP 6.1.6.1 Limit

Page 13


DIABLO CANYON POWER PLANTANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002

TABLE ICGASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES

Units Annual Total


1. Total release Ci 1.68E+1

2. Average release rate for period pCi/sec 5.33E-1

3. Percent of technical specification limit' 1.56E-3

B. lodines

1. Total iodine-131 Ci 2.37E-3

2. Average release rate for period pCi/sec 7.50E-5

3. Percent of technical specification limit' % 4.19E-4

C. Particulates

1. Particulates with half-lives >8 days Ci 9.17E-4

2. Average release rate for period pCi/sec 2.91 E-5

3. Percent of technical specification limit' 1.08E-5

4. Gross alpha radioactivity Ci 5.81 E-7

D. Tritium


2. Average release rate for period pCi/sec 9.74E+O

3. Percent of technical specification limit' _ 2.29E-5

i

'RECP 6.1.6.1 Limit

Page 14




TABLE 2A

GASEOUS EFFLUENTS - GROUND LEVEL RELEASES

First Quarter Second Quarter

Nuclides Released Units Continuous Batch Mode Continuous Batch Model~ oe BthMode Mode Bac

1. Fission gases

argon-41 Ci 2.54E+O 4.44E-2 MDA 1.13E+O

krypton-85 Ci MDA 2.OOE-1 MDA 2.40E-1

krypton-85m Ci MDA MDA MDA MDA

krypt6n-87 Ci MDA MDA MDA MDA

krypton-88 Ci MDA MDA MDA - MDA

xenon-131m Ci MDA MDA MDA 7.52E-4

xenon-133 Ci 8.01E+O 2.23E-2,. MDA'. 1.47E-1

xenon-1 33m Ci MDA MDA - MDA 7.64E4 -

xenon-135 Ci MDA 4.30E-5 MDA 7.78E-5

xenon-135m Ci MDA MDA MDA MDA

xenon-I 38 Ci MDA MDA MDA MDA

TOTAL FOR PERIOD Ci 1.06E+1 2.67E-1 MDA I.52E+O

2. lodines

iodine-131 Ci 1.71 E-4

iodine-I 33 Ci MDA

iodine-135 Ci MDA

TOTAL FOR PERIOD Ci 1.71 E-4

1.76E-5

MDA

MDA

1.76E-5


Page 15




TABLE 2A (Continued)


Continuous Mode

Nuclides Released |U Units First Quarter Second Quarter

3. Particulates

barium-140 Ci MDA MDA

cesium-134 Ci MDA MDA

cesium-1 37 Ci 2.97E-4 MDA

cerium-141 Ci MDA MDA


chrornium-51 .. Ci MDA MDA

coba t-57 Ci MDA MDA

cobalt-58 Ci 2.06E-4 2.65E-4

cobalt-60 Ci MDA MDA

iron-59 Ci MDA MDA

lanthanum-140 Ci MDA MDA

manganese-54 Ci MDA MDA

molybdenum-99 Ci MDA MDA

ruthenium-103 Ci MDA MDA

strontium-89 Ci MDA MDA


zinc-65 Ci MDA MDA

zirconium-95 Ci MDA MDA

TOTAL FOR PERIOD Ci 5.04E-4 2.65E-4

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume).This note applies to all tables.

Page 16



ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002TABLE12B


--- - -II

iIi

i

iiI

i

Third Quarter Fourth Quarter

Nuclides Released Units I Mode B Model Mode B

1. Fission gases

argon-41 Ci MDA 2.22E-2 8.73E-1 1.84E-2

krypton-85 Ci MDA 5.53E-2 MDA 2.76E-1

krypton-85m Ci MDA MDA MDA MDA

krypton-87 Ci MDA MDA MDA MDA

krypton-88 Ci MDA, MDA -MDA MDA

xenon 131m r Ci MDA 1.12E-2 ' 3 MDA 1.04E-2

xenon-133 Ci MDA ' 3.98E-1I MDA 2.51 E+0

xenon-133m Ci MDA 1.14E-3 t MDA 1.91E-4

xenon-135 Ci 2.71 E-1 1.04E-4 5.39E-5. 3.21 E-2

xenon-135m Ci MDA MDA .> MDA MDA

xenon-1 38 Ci MDA MDA MDA MDA

TOTAL FOR PERIOD Ci 2.71 E-1 4.88E-1 8.74E-1 2.84E+O

2. lodines

iodine-1 31 Ci 1.95E-5

iodine-133 Ci 3.55E-5

iodine-135 Ci MDA

TOTAL FOR PERIOD Ci 5.50E-5

2.02E-3

1.06E-4

MDA

2:12E-3

MDA = Less than the "a posteriori minimum detectable activity (microcuries per unitmass or volume). This note applies to all tables.

Page 17



ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002TABLE 2B (Continued)


Continuous Mode

INuclides Released Units | Third Quarter | Fourth Quarter

3. Particulates

barium-140 Ci MDA MDA

cesium- 134 Ci MDA MDA

cesium-1 37 Ci MIDA 1.25E-4


cerium- 144 Ci MDA MDA

chromium-51 Ci MDA . MDA

cobalt-5T, t- Ci MDA r--: MDA;

cobalt-58 - Ci 2.35E-5 'MDA '

cobalt-60 !Ci MDA - MDAl

iron-59 Ci MDA MDA

lanthanum-140 Ci 'MDA MDA

manganese-54 Ci MDA MDA

molybdenum-99 Ci MDA MDA

ruthenium-I 03 Ci MDA MDA


strontiun-90 Ci MDA MDA

zinc-65 Ci MDA MDA

zirconium-95 Ci MDA MDA

TOTAL FOR PERIOD Ci 2.35E-5 1.25E-4

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume).This note applies to all tables.

Page 18



ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002TABLE 3

GASEOUS EFFLUENTS - LOWER LIMITS OF DETECTION

I

1. Fission gases

argon-41 pCi/ml 1.52E-8 1.52E-8 1.52E-8

krypton-85 pCi/ml 2.37E-6 2.37E-6 2.37E-6

krypton-85m pCi/ml 8.22E-9 8.22E-9 8.22E-9

krypton-87 pCVMI 2.90E-8 2.90E-8 2.90E-8

krypton-88 pICi/ml 2 88E-8 2.88E-8 2.88E-8

xenon- 131m pi ILCi/mIl 2.79E-7 2.79E-7 2.79E-7.

xenon-133 , pCi/ml 1.99E-8 1 .99E-8 I.99E28 -

xenoh-133m P i C/mi 5.77E-8 5.77E-8 - n 5.77E-8 A

xenon-1 35 ; pCi/ml 6.84E-9 6.84E-9 . 6.84E-b *

xenon-1 35m RCi/ml 2.54E-7 2.54E-7 . 2.54E-7

xenon-1 38 PC/mi' 7.37E-7 7.37E-7 7.37E-7

2. Tritium

hydrogen-3 pCi/ml 4.18E-9 4.18E-9 N/A

3. lodines

iodine-131 pCi/mi 4.87E-13 N/A

iodine-133 pC/mI 7.36E-13 N/A

iodine-135 pCi/ml 3.26E-12 N/A

Page 19




TABLE 3 (Continued)GASEOUS EFFLUENTS - LOWER LIMITS OF DETECTION

Nuclide Units Continuous Mode

4. Particulates

barium-140 pCi/ml 1.96E-12

cesium-1 34 PCiml 3.80E-1 3

cesium-137 PCi/ml 6.63E-13

cerium-141 pCi/ml 5.17E-13

cerium-144 PCi/ml 2.15E-12

chromium-51 pCi/ml 2.76E-12

cobalt-57 pCivmr n 2.44E-13

cobalt-58 PCi/mi 5.09E-13 3

cobalt-60 iquCiml 6.74EE-3j ,.

, 2 . iron-59 , pCi/mi 1.26E-12

lanthanum-140 PCi/ml 6.38E-13

manganese-54 [pCi/ml 4.31 E-1 3

molybdenum-99 pCilml 2.34E-13

ruthenium-103 pCVmi 4.11 E-13

strontium-89 -PCvml 2.1OE-14

strontium-90 pCi/ml 6.52E-15

zinc-65 pCi/ml 1.22E-12

zirconium-95 pCI/ml 7.32E-13

gross alpha pCi/ml 2.35E-15

Page 20



ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002TABLE 4A

LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES

Units First Second Est TotalQuarter Quarter Error %

A. Fission and activation products

1. Total release (not including tritium, gases, alpha) Ci 2.26E-2 3.82E-2 24% |

2. Average diluted concentration during period pCi/ml 8.29E-12 1.73E-1 1

3. Percent of applicable limit' 1.67E-4 2.73E-4

B. Tritium

I 1. Total release I Ci I 1.84E+2 | 4.11E+2 | 13% l

2. Average diluted concentration during period pCimI 6.75E-8 1.86E-7

! I13. Percnt of applicable limt'E j2 6 5E3jI

II V I. IIII

IC. Dissolved and entrained passes

1. Total release MDA 3.79E-4 24%

2. Average diluted concentration during period pCI/ml MDA 1.71 E-13

3. Percent of applicable limit' % MDA 8.56E-8

D. Gross Alpha

1. Total release Ci MDA 2.36E-5 | 61%

a I I I

E. Volume of waste release (prior to dilution) - I liters I 9.27E+7 I 8.53E+7 I 5%

I I

F. IVolume of circulating saltwater used duringrelease periods

I liters I2.72E+12 12.21 E+ 121 7%.


1 RECP 6 1.3.1 Limit

Page 21




TABLE14B


UIts Third | Fourth Est.TotalQuarter | Quarter Error %


1. Total release (not including tritium, gases, alpha) Ci 2.98E-2 4.17E-2 0

2. Average diluted concentration during period piCilml I 2E 1 2.00E1 1

3. Percent of applicable limit' % 2.18E-4 8.03E-4

B. Tritium

1. Total release Ci 2.29E+2 5.50E+2 0

Average diluted concentration during period iCicml 8.59E 8 2.64E-7

i3. Percent of aplicable limit % 859E-3 2.64E-2

C. Dissolved and entrained gasses

1. Total release Ci 2.34E-3 9.83E-3 -

2. Average diluted concentration during'period - -iCi/ml 8.80E-13 4.72E-12

3. Percent of applicable limit'- 4.40E-7 2.36E-6

D. Gross Alpha

1. Total release Ci MDA MDA 61%

II.

I

4

_ e I

E. kolume of waste release (prior to dilution) -1liters I8.40E+7 I8.05E-+7I 5%

. IF. IVolume of circulating saltwater used during

release periodsls 2.66E+12 2.08E+12

I

.MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unitmass or volume). This note applies to all tables.

1 RECP 6.1.3.1 Limit

Page 22


DIABLO CANYON POWER PLANT w I

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002TABLE 4C


I Units [ Annual Total


1. Total release (not including tritium, gases, alpha) Ci 1.32E-1

2. Average diluted concentration during period jLCi/ml 1.37E-11

3. Percent of applicable limit' 3 42E-4

B. Tritium


2. Average diluted concentration during period laCi/ml 1.42E-7

-3. Percent of applicable limit, , 1.42E-2

- I'C. Dissoivedaand entrained gassesIe --. r'

1. Total release - Ci ' 1.26E-2 .

2. Average diluted concentration during period pCi/ml I.30E-12

3. Percent of applicable limit' % 6.49E-7

D. Gross Alpha

1. Total release Ci 2.36E-5

Is

E. jVolume of waste release (prior to dilution) _

F. IVolume of circulating saltwater used during. release periods

I liters I 3.43E+8I

I I

I liters I 9.68E+12 IMDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume).This note applies to all tables.

'RECP 6.1.3.1 Limit

Page 23


DIABLO CANYON POWER PLANT -_


TABLE 5ALIQUID EFFLUENTS - NUCLIDES RELEASED


Nuclides Released Units Continuous Continuous Batch ModeMoe BthMode Mode _____

antimony-I 22 Ci MDA MDA MDA 3.28E-5

antimony -124 Cl MDA MDA MDA 1.91 E-4

antimony-125 Ci MDA 5.24E-3

barium-140 Ci MDA MDA MDA MDA

beryllium-7 Ci MDA MDA MDA MDA

bromine-82 Ci MDA MDA MDA MDA

cerium-141 Ci MDA MDA MDA MDA

cerium-143 Ci MDA MDA MDA MDA

cerium-144 Ci MDA MDA -MDA MDA

cesium-134 . Ci .. MDA - 1.50E-4 .MDA 4.95E-4.

cesium-136 * Ci , ,MDA, MDA ,MDA ,-MDA,-

cesium-137 Ci 2.21 E-4 . 4.78E-4 MDA 1.25E-3

cesium-I 38 Ci -MDA MDA MDA MDA

chromium-51 Ci MDA MDA MDA 1.39E-3

cobalt-57 Ci MDA 6.34E-5 MDA 1.15E-4

cobalt-58 Ci MDA- 3 25E-3 1.29E-4 1.03E-2


iron-55 Ci MDA 8.20E-3 MDA 7.51 E-3

iron-59 Ci MDA MDA MDA- 4.98E-5

lanthanum-140 Ci MDA MDA MDA MDA

-I . -

I

4 .

I

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass orvolume). This note applies to all tables

Page 24




TABLE 5A (CONTINUED)

LIQUID EFFLUENTS - NUCLIDES RELEASED

T

First Quarter -Second Quarter

Nuclides Released Units Continuous Continuous Batch ModeMoe BthMode Mode _____


manganese-54 Cl MDA 2.24E-4 MDA 3.24E-4

manganese-56 Cl MDA MDA MDA MDA

molybdenum-99 Ci MDA 1.03E-5 MDA 1.97E-5

niobium-95 Ci MDA MDA MDA 8.53E-5

neodymium- 147 Ci MDA MDA MDA MDA

rubidium-89 Ci MDA MDA MDA MDA

ruthenium-103 Ci MDA MDA MDA MDA

silveri'110m Ci MDA -9.11 E-6 ..MDA 9.64E-5'

sodium-24 - Ci MDA MDA MDA MDA

strontium-89 - Ci MDA MDA MDA vMDA:

strontium-90 Ci MDA 5.81 E-5 MDA MDA

strontium-91 Ci I MDA MDA MDA MDA,

strontium-92 Ci MDA 1.OOE-5 MDA 8.51 E-6

tellurium-129m Ci MDA MDA MDA MDA

tellurium-131 Ci MDA MDA MDA MDA


tin-113 Ci MDA MDA MDA MDA

I - -

I

MDA = Less than the "a posterior" minimum detectable activity (microcuries per unit mass orvolume). This note applies to all tables.

Page 25


DIABLO CANYON POWER PLANT ;


TABLE 5A (CONTINUED)



Nuclides Released Units Continuous Batch Mode Continuous Batch ModeMode _ _ _ __ Mode _ _ _ _ _

tin-1 17m Ci MDA MDA MDA MDA

tungsten- 87 Ci MDA MDA MDA MDA

zinc-65 Ci MDA 1.72E-3 MDA 2.64E-3

zirconium-95 Ci MDA MDA MDA MDA

iodine-i31 Ci 5.01E-4 4 65E-4 MDA 1.27E-4

iodine-1 32 Ci MDA MDA MDA MDA

iodine-133 Ci MDA 1.67E-5 MDA 5.88E-6

iodine-134 > Ci MDA MDA * - MDA MDA

iodine-135 i , Ci^ 5 MDA MDA i.- 3MDA MDA

TOTAL FOR PERIOD Ci 7.22E-4' 2.18E-2:T :1:29E-4 v3.81 E-2

DISSOLVED AND ENTRAINED GASES

xenon-133 Ci MDA MDA MDA 3.79E-4

xenon-1 33m Ci MDA MDA MDA MDA

xenon-I 35 Ci MDA MDA MDA MDA




TOTAL FOR PERIOD Ci MDA MDA MDA 3.79E-4

I

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass orvolume) This note applies to all tables.

Page 26




TABLE 5B



Continuous CniuuNuclides Released Units Batch Mode Conu Batch Mode

______ ModeA M Mode | MDA

antimony-122 Ci MDA MDA MDA MDA

antimony-I124 Ci MDA MDA MDA -MDA

antimony-1 25 Ci MDA 1.59E-3 MDA 2.02E-3

barium-140 Ci MDA MDA MDA MDA

beryllium-7 Ci MDA MDA MDA MDA

bromine-82 Ci MDA MDA MDA MDA

cerium-141 Ci MDA MDA MDA - MDA

cerium-143 . Ci MDA MDA MDA MDA

ceriurn-14"" Ci* MDA-- MDA- MDA MDA".

cesium-134 - . -Ci- MDA -5.96E-4 1:31E-3 7.71E-4

cesium-136t Ci' MDA MDA- .* MDA v 7.04E-7 .

cesium-137 Ci MDA 1.38E-3 1.98E-3 1.56E-3

cesium-1 38 Ci; MDA MDA MDA MDA

chromium-51 Ci' MDA 2.61 E-4 MDA 3.55E-5


cobalt-58 Ci MDA 1.OOE-2 MDA 8.94E-3

cobalt-60 Ci MDA - 5.33E-3 MDA 6.56E-3

iron-55 Ci MDA 6.17E-3 MDA 7.97E-3

iron-59 Ci MDA MDA MDA MDA


i

. - I � ;.

-1... Z

.- I .. I

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass orvolume). This note applies to all tables.

Page 27

- - e, - 4r EnclosurePG&E Letter DCL-03-049


ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002TABLE 5B (CONTINUED)

LIQUID EFFLUENTS -NUCLIDES RELEASED

Third Quarter Fourth-Quarter

I- � - I4 -

� -, I f

q- "

I 4

Continuous CniuuNuclides Released Units Batch Mode Conu Batch Mode

_____ Mode ______ Mode _____

lanthanum-I 42 Ci MDA MDA MDA MDA

manganese-54 Ci MDA 2.64E-4 MDA 1.64E-4

manganese-56 Ci MDA MDA MDA - MDA

molybdenum-99 Ci MDA 4.96E-5 MDA 2.08E-4

niobium-95 Ci MDA 3 63E-4 MDA 4.15E-5

neodymium-147 Ci MDA MDA MDA MDA

rubidium-89 Ci MDA MDA MDA MDA

ruthenium-103 4- . Ci MDA 'MDA MDA MDA .

silver-IOm 'MDAC0MDA 5.73E-5 ,MDA2E-4

sodiun,-24' Ci' MDA MDA . MDA

strontiu'm-89 Me ! . Ch MD MDA MDA

strontium-90 Ci MDA MDA MDA MDA

strontium-91 - Ci MDA MDA nMDA MDA

strontium-92 Ci MDA 5.09E-6 MDA 4.53E-6'

tellurium-129m Ci MDA MDA MDA MDA


telluum-I 32 Ci MDA MDA MDA MDA

tin-1 13 Ci MDA 4.50E-6 MDA MDA

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass orvolume). This note applies to all tables.

Page 28

S ^ *J EnclosurePG&E Letter DCL-03-049



TABLE -B (CONTINUED)

LIQUID EFFLUENTS -`NUCLIDES RELEASED


II- . -71f

I t , "'I- 'i 11-4.

-. -f� Z ,, -

.1 -ii�_ � .

Nucides Released Units Continuous Batch Mode Continuous Batch Mode

tin-1 17m Ci MDA MDA MDA MDA

tungsten-I 87 Ci MDA MDA MDA MDA

zinc-65 Ci MDA 2.47E-3 MDA 1.79E-3

zirconium-95 Ci MDA 1.26E-4 MDA MDA

iodine-131 Ci 8.88E-5 7.12E-4 3.85E-3 3.76E-3

iodine-1 32 Ci MDA MDA MDA 9.89E-7

iodine-133 Ci 1.61 E-4 3.63E-5 7.26E-5 4.49E-4

iodine-134 Ci MDA HMDA MDA MDA

iodie-I35 Ci '-MDA - MD'Ai MDA 7l5.43E-6

TOTAL FOR PERIOD Ci 250E-4 - 2.95E:2; 7.22E-3 3.45E-2

DISSOLVED AND ENTRAINEDGASES

xenon-133 Ci MDA 2.34E-3 MDA 9.78E-3

xenon-133m Ci MDA MDA MDA MDA

xenon-I 35 Ci MDA MDA 3.63E-5 2.09E-5


krypton-87 Ci MDA MDA' MDA MDA

krypton-88 Ci MDA MDA - MDA MDA

TOTAL FOR PERIOD Ci MDA 2.34E-3 3.63E-5 9.80E-3

If ,I ,I1� , '-

I

.,.I

i

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or

volume). This note applies to all tables.

Page 29

EnclosureI I PG&E Letter DCL-03-049



TABLE 6LIQUID EFFLUENTS - LOWER LIMITS OF DETECTION

Nuclide Units LLD

antimony-122 ,uCi/ml 1.14E-7

antimony-124 pCi/ml 6.70E-8

antimony-1 25 pCi/ml 2.72E-7

barium-140 pCi/ml 3.20E-7

beryllium-7 pCi/ml 7.75E-7

bromine-82 pCi/mI 1.18E-7




cesium-134 pCi/ml 6.74E-8


cesium-137 PCi/mI 1 .54E-7


chromium-51 pCi/ml 6.48E-7

cobalt-57 ,uCi/ml 5.19E-8

cobalt-58 pCi/ml 1.48E-7

cobalt-60 pCi/ml 1.21 E-7

iron-55 pCVmI 3.OOE-7

iron-59 pCi/ml 2.02E-7

lanthanum-140 pCi/ml 1.98E-7

lanthanum-142 pCi/ml 3.75E-7

manganese-54 'tCi/ml 1.16E-7

manganese-56 pCi/mi 4.33E-7

molybdenum-99 / pCVml 6.43E-8

niobium-95 pCi/ml 1.08E-7

Page 30




TABLE 6 (CONTINUED)

LIQUID EFFLUENTS - LOWER LIMITS OF DETECTION

Nuclide Units LLD

neodymium-147 pCilml 2.83E-7

rubidium-89 pCi/ml 5.58E-6

ruthenium-I 03 pCi/ml 1.01 E-7

silver-11 Om 11Ci/ml 8 70E-8

sodium-24 pCi/ml 1.54E-7

strontium-89 ,Ci/ml 4.95E-8

strontium-90 PiCUmi 1.78E-8

strontium-91 pCiml 1.45E-7

strontium-92 pCiml 2.56E-7

tellurium-129m pCi/ml 3.35E-6

tellurium-131.. pCi/ ,. -m 4.58E-7. B

tin- 117mn ,pCi/ml: 5.55E-8 *

tungsten-187 pCi/ml 3.40E-7

zinc-65 : - pCimI 2.51 E-7

zirconium-95 pCiml 1 .38E-7

gross alpha 1iCirl 9.12E-8

hydrogen-3 ,Ci/ml 3.97E-6

I f --

i - I

:,I .

I , Ak,.

I

I

Page 31




TABLE 6 (CONTINUED)

LIQUID EFFLUENTS - LOWER LIMITS OF DETECTION

Nucide Units LLD

iodine-I 31 ,uCi/ml 8.13E-8

iodine-132- ,pCi/ml 1.27E-7

iodine-133 pCiml 1.18E-7

iodine-1 34 pCvml 2.76E-7

iodine-135 ,uCi/ml 5.90E-7

xenon-1 33 [tCl/ml 7.54E-8

xenon-1 33m -pCi/ml 2.01E-7

xenon-1 35 pCi/ml 2.40E-8

krypton-85 pCi/ml 1.18E-5

krypton-87 pCi/ml 1.01E-7

krypton-88 , . . pCimI , 8.92E-8,.

. .b I.. t

I I

P~ I y-

j ,rF h

I >i - " X , t

I 'w- . ", 1 '

II I

twz

t

_ 11 .4.

II II _ .-

; . I

1. I

.. I

4I

Page 32


IX. Solid Radwaste Shipments

Page 33


IX. Solid Radwaste Shipments

Solid Waste and Irradiated Fuel Shipment

A. Solid Waste Shipped Off-site for Burial or Disposal (Not irradiated fuel)

1. Type of Waste Unit 12 Month Period Est. Total Error, %

a. ,Spent Resins, Filter Sludges, m3 1.01 E+01Evaporator Bottoms, etc. Ci 3.23E+2 8.99E+O

b. Dry Compressible Waste, m3 7.58E+or' Contaminated Equipment, etc. Ci 1.27E+0 -9.0E+0

c. Irradiated Components, Control m3 0.OOE+ORods, etc. Ci O.OOE+0 O.00E+0-

d. Other m3 0.OOE+O -

ICi 0.OOE+0 O.OOE+0

2. Estimate of Major Nuclide Composition (by type of waste)

a.

b.

Co-60 % 3.50E+01

Ni-63 % 1.60E+01

Fe-55. % 1.30E+01

Co-58- 8.2E+00

Nb-95 % 4.6E+00

Fe-55 __3.02E+01

Co-60 % 1.84E+01

Co-58 1.49E+01

Mn-54 % 1.42E+01

H-3 % 8.24E+00

Ni-63 % 3.75E+00

Nb-95 _ _% -- 2.47E+00

C-14 % - 1.67E+00

Zr-95 % 1 .58E+00

c. I Not Applicable .1 % I N/A I. I --

d. I Not Applicable I % I N/A I. I

Page 34

Enclosure- PG&E Letter DCL-03-049

Solid Waste and Irradiated Fuel Shipment (Continued)

3. Solid Waste Disposition

Number of Shipments -Mode of DestinationTransportation

3 Truck BARNWELL, SC

19 Truck CLIVE, UT

4. Supplemental Information Required by former TS 6.9.1.6

Solidification Type of Container Number of 10 CFR 61 Shipping TypeAgent Containers Waste

___ ___ __ Class -

Cement Strong, Tight 1 C - Type APolymer Strong, Tight -1 B - Type B

None Type A, HIC i -C Type BNone Strong, Tight 21 - AU LSA

B. Irradiated Fuel Shipments (Disposition)

Page 35


X. Radiation Dose Due to Gaseous and Liquid Effluents

Radiation Doses

A. Radiation doses from radioactive liquid effluents

The radiation dose contributions due to releases of radioactive liquid effluents to thetotal body and each individual organ for-the maximum exposed adult have beencalculated in accordance with the methodology in the ODCP. Dose contributionslisted in Table 7 show conformance to RECP 6.1.4.1.

B. -Radiation doses from radioactive'gaseous effluents

The radiation dose contributions due to radioactive gaseous effluents at the siteboundary for the land sectors have been calculated in accordance with thecalculation methodology in the ODCP. Each unit's dose contribution has beencalculated separately. The latest five-year historical average meteorology -conditions were used in these calculations. In addition to the site boundary doses,the dose to an individual (critical receptor) due to radioiodines, tritium, and.-particulates released in gaseous effluents with half-lives greater than eight days isdetermined in accordance with the methodology in the ODCP based on themethodology described in NUREG-0133. Dose contributions listed in Table'8,which represents the maximum dose for age groups, organs, and geographiclocations for the report period, show conformance to RECP 6.1.6.1, 6.1.7.1, and6.1 .8.1 .

C. Radiation Doses from Direct Radiation (Line-of-Sight Plus Sky-Shine) -Closest Site Boundary (800 m)

For the report period, the radiation dose is evaluated to be 6.58E-3 M-R due to thepresence of radioactive waste containers outside of plant buildings and the storageof contaminated tools and equipment inside plant buildings.

D. Radiation Doses from Chemistry Laboratory Radioactive Gaseous Effluents -Closest Site Boundary (800m)

The radiation dose due to chemistry laboratory radioactive gaseous effluents for thereport period is evaluated to be 9.90E-6 mR.

E. Radiation Doses from lPost accident Sampling System Radioactive GaseousEffluents - Closest Site Boundary (800m)

The radiation doses due to post accident sampling system radioactive gaseouseffluents for the report period is evaluated to be 7.59E-6 mR.

Page 36


F. 40 CFR 190 Considerations

The release of radioactivity in liquid and gaseous effluents resulted in doses that aresmall percentages of the TS limits as shown in Tables 9 and 10. This, coupled withthe fact that there are no other uranium fuel cycle sources within eight kilometers of-the DCPP, shows conformance to 40 CFR 190.

G. Radiation Doses from Radioactive Liquid And Gaseous Effluents to Membersof the Public Due To Their Activities Inside The Site Boundary

1. Liquid Effluents

The radiation dose to members of the public within the site boundary-due to therelease of radioactive liquid effluents is negligible because the discharge piping forliquid radwaste is mostly' imbedded in concrete, located in remote or inaccessibleareas, or is underground. In addition,;the quantity of radioactivity released was verylow.


The radiation dose to members of the -public within the site boundary due to therelease of radioactive gaseous effluents-are listed in Table 11.

Page 37

- -1



ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002TABLE 7

RADIATION DOSE DUE TO THE RELEASE OF RADIOACTIVE LIQUID EFFLUENTS

millirem. . r. I

First Quarter Second Quarter Third Quarter Fourth Quarter Annual Total

Total Body 2.74E-04 5.05E-04 3.64E-04 4.84E-04 1.63E-03

Bone 3.70E-04 5.33E-04 - 3.72E-04 5.38E-04 1.81E-03

Liver 6.33E-04 1.09E-03 7.98E-04 1.OOE-03 3.52E-03

Thyroid 6.68E-05 5.35E-05 5.85E-05 7.07E-04 8.85E-04

Kidney 3.21 E-04 6.03E-04 4.55E-04 5.29E-04 - 1.91 E-03

Lung 1.OOE-04 1.52E-04 8.25E-05 1.85E-04 5.19E-04

G.I. LLI 5.13E-04 1.08E-03 9.34E-04 8.90E-04 3.42E-03

f ,

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Page 38


, * t I . . ,, I

DIABLO CANYON POWERkPLANT

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002TABLE 8A,

RADIATION DOSE1 DUE TO THE RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS (UNIT 1)

-

First Quarter Second Quarter Third Quarter Fourth QuarterDose . , I - Dose I Dose ,I Dose

.. _ . .

IAnnual Total

Dosev _ _ __ _ _ _ _ _ _ - _ _ _

Nearest Residence-NNW _ _ e_,_-

I 13±, IT2.3 ICritical Receptor (Highest Organ) mrem 2.83E-4 | 1.69E-3 | 6.26E-4 | - 1.59E-3 | 4.19E-3

| First Quarter Second Quarter Third Quarter Fourth Quarter Annual TotalDose- Dose Dose Dose 4 Dose

Nearest Vegetable Garden - ESE __ ._,

IPT Recepto (ie Or 1 | 1.59E-4 | 1| 3.

ICritical Receptor (Highest Organ) T ~mrem j 15E4 I .1.02E-3 I 3.51 E-4 8.94E-4 2.42E-3

I, e

11 Page 3'9

. , I so-:%

I L


DIABLO CANYON POItRPANT


TABLE 8B

RADIATION DOSE1 DUE TO THE RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS (UNIT 2)

First Quarter | Second Quarter I Third Quarter Fourth QuarterDose Dose Dose I Dose

Annual IotalDose

Nearest Residence-NNW . , . ICrticalReceptor (Highest Organ) mrem 5.04E-4 2.892.65E4 1.09E-3 | 2.15E-3

First Quarter Second Quarter Third Quarter Fourth Quarter Annual Totall Dose Dose | Dose | Dose | Dose

Nearest Vegetable Garden - ESE

mrem 1.63E-4 1 2.55E-3ICritical Receptor (Highest Organ) I rm I 4.48E-4 j 163- 1.67E-4 25E3 3.33E-3

." i " � , , 1. -� -,, Z-.,. : I �

Page 40" I ~,.

ii

ii


Notes for Tables 8a and 8b

1. This represents the maximum dose of age groups, organs, and geographic locationsfor the quarter and the year.

2. Radioiodines,- radioactive material in particulate form, and radionuclides other thannoble gases with half-lives greater than eight days.

3. Thelinhalation and ground plane pathways are included for this location.

4. The inhalation, ground plane'and vegetable pathways are included for this location.An occupancy factor of 0.5 was used for the inhalation and ground plane pathways.The teen age group had the highest calculated dose for this location.

Page 41



TABLE 9

PERCENT OF TECHNICAL SPECIFICATION LIMITS' FOR RADIOACTIVE LIQUID EFFLUENTS

- PercentIa. .

First Quarter Second Quarter IThird Quarter Fourth Quarter Annual Total

Total Body ~ 1.83E-02 3.37E-02 2.42E-02 3.22E-02 5.42E-02

Bone 7.39E-03 1 .07E-02 7.44E-03 1 .08E-02 1.81 E-02

Liver 1.27E-02 2.18E-02 1.60E-02 2.01 E-02 3.52E-02

Thyroid 1.34E-03 1.07E-03 1.17E-03 1.41E-02 - 8.85E-03

Kidney 6.43E-03 1.21 E-02 9.11 E-03 1.06E-02 1.91 E-02

Lung 2.01 E-03 3.04E-03 1.65E-03 3.70E-03 5.19E-03

G.I. LLI 1.03E-02 2.16E-02 . 1.87E-02 1.78E-02 3.42E-02

NOTE:

'RECP 6.1.4.1 * t 1

{~ ..

.~ . .

I -

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Page 42

I EnclosurePG&E Letter DCL-03-049


TABLE1 OAPERCENT OF TECHNICAL SPECIFICATION LIMITS1 FOR RADIOACTIVE GASEOUS EFFLUENTS (UNIT 1)

First Quarter %of TS Limit I

Second Quarter I Third Quarter % Fourth Quarter % Annual Total %% of TS Limit I of TS Limit I of TS Limit | of TS Limit *1

- . , -::-

Site Boundary _ JNoble GasGamma Air Dose mrad 8.55E-2 3.63E-2 4.05E-4 2.10E-2 7.16E-2

,Beta Air Dose mrad 1.74E-2 v u7.36E-3 7.14E - 7.75E - 1 63E-

First Quarter % Second Quarter Third Quarter % Fourth Quarter %I Annual Total %I of TS Limit , % ofTS Limit I of TS Limit of TS Limit of TS Limit

Nearest Residence - NNWI IPI T I .. .

Critical Receptor (Highest Organ) I mrem 3.78E-3. . 2.25E-2 8.34E-3 2.12E-2 2.79E-2

First Quarter % Second Quarter Third Quarter % Fourth Quarter %°Annual Total %of TS Limit % of TS Limit of TS Limit of TS Limit of TS Limit l

Nearest Vegetable Garden - ESE I.P. ., l T l . lCritical Receptor (Highest Organ) mrem ' | 2.12E-3 | 1.36E-2 | 4.68E-3 I 1.19E-2 I 1.61 E-2

NOTE:'RECP 6.1.6.1, 6.1.7.1 and 6.1.8.1

IPige 43

P~&~ ejterEnclosurePG&E Letter DCL-03-049


TABLE 1OBPERCENT OF TECHNICAL SPECIFICATION LIMITS1 FOR RADIOACTIVE GASEOUS EFFLUENTS (UNIT 2)

;Y. _ . . . .

First Quarter %I nf T Limit I

Second Quarter Third Quarter % Fourth Quarter %'0o of TS Limit of TS Limit I of TS Limit

I Annual Total %of TS Limit

._ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ .1 - ..._.

Site Boundary ___

Noble Gas I . . . iGamma Air Dose mrad 8.37E-3 4.91E-4 2.63E-3 1.12E-2 1.14E-2

Beta Air Dose I mrad * 1 .28E-2 - - 1.74E-4 2.16E-3 3.04E-3 9.09E-3'

First Quarter % Second Quarter Third Quarter % Fourth Quarter % Annual Total %

F ofTS Limit- 1 -% ofTS Limit I of TS Limit | of TS Limit of TS Limit

Nearest Residence-NNW

Critical Receptor (Highest Organ) | mrem | 6.72E-3 | 3.86E-3 | 3.53E-3 | 1.45E-2 | 1.43E-2

NOTE:'RECP 6.1.6.1, 6.1.7.1 and 6.1.8.1

I I'.z- , , . , r

, Page 44,. 1 . . "

." " .1 I I - -, - I I .1 ". - . -7,


DIABLO CANYON POWERP.ILANTANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2002

TABLE 11ARADIATION DOSE DUE TO-RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS,

I , FIRST QUARTER, 2002ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUPS)

- External Dose Internal Dose

Noble Gas lodines, Particulates, and Tritium 1

Locations Closest posure TDin Whole Body Skin Ground Plane InhalationSpecific Activity (Sectors) (meters) (Hours)

Police at Shooting Range SE 700 52.0 1.62E-4 2.51E-4 1.90E-4 1.48E-4

Tour Participants(a) Simulator Bldg. S 310 1.00 .2.12E-6 3.30E-6 1.32E-6 1.94E-6

(b) Bio Lab SSE 460 1.50 3.18E-6 4.94E-6 3.OOE-6 2.90E-6

(c) Overlook E 210 . 0.25 1.48E-6 2.29E-6 4.31 E-7 1.35E-6'American Indians NW 200 24.0 5.86E-4 9.1OE-4 1.97E-4 5 35E-4at Burial Grounds NNW 200 24.0 4.1OE-4 6.37E-4 1.12E-4 3.74E-4Ranch Hands driving NW 250 0.25 . 4.09E-6 6.35E-6 1.44E-6 - 3.73E-6cattle around site NNW 350 0.25 1.57E-6 2.43E-6 4.87E-7 1.43E-6

N 320 0.25 ,; .04E-6 1.62E-6 2.45E-7 9 49E-7

NNE 450 0.25 6 ; - 4.07E-7 6.32E-7 1.06E-7 3.71E-7, NE 630 0.25 2 .' 2.07E-7 3.21 E-7 6.49E-8 1.89E-7

NOTE: All doses are in mrem.* IPage 45

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- TABLE iI1B -RADIATION DOSE DUE TO RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS

SECOND QUARTER, 2002ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUPS)

External Dose | Internal Dose

Noble Gas lodines, Particulates, and Tritium - .

Exposure Exposure EoueTmSpecific Activity Location Closest Dist W(Hours) hole Body Skin Ground Plane Inhalation

(Sectors) (meters), 1ors

Police at Shooting Range - SE 700 52.0 6.42E-5 9.59E-5 6.09E-6 4.67E-4

Tour Participants(a) Simulator Bldg. S 310 1.00 8.42E-7 1.26E-6 4.23E-8 6.13E-6

(b) Bio Lab SSE 460 1.50 - 1.26E-6 1.89E-6 9.61 E-8 9.20E-6

(c) Overlook E 210 0.25 5.86E-7 8.76E-7 1.38E-8 4.27E-6

American Indians NW 200 24.0 2.32E-4 3.47E-4 6.29E-6 1.69E-3

at Burial Grounds NNW 200 24.0 1.63E-4 2.43E-4 '3.59E-6 1.18E-3

Ranch Hands driving NW 250 0.25 1.62E-6 2.43E-6 4.62E-8 1.18E-5

'cattle around site NNW, 350 0 25-, ., 6.21 E-7 9.29E-7 1.56E-8 . 4.52E-6

N 320 0.25 -4.13E-7 6.17E-7 7.85E-9 3.01E-6

NNE 450 0.25 w 1.61E-7 2.41E-7 3.38E-9 1.18E-6

NE 630 0.25 - 8.21 E-8 1.23E-7 2.08E-9 5.98E-7

NOTE: All doses are in mrem.F~aje 46

I I". I -j - - . -

1, . .. .. - .", "

- ,,' -

Dr

Enclosure&E Letter DCL-03-049

DIABLO CANYON POWER PLANTANNUAL RADIOACTIVE EFFLUENTARELEASE REPORT 2002

.TABLE i1CRADIATION DOSE DUE TO RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS

THIRD QUARTER, 2002ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUP

:1 -

S)

I . .I , External Dose 1 internal Dose

Noble Gas lodines, Particulates, and Tritium I- .. . .. - -. _ _ , -

SpecificActivity Locson Exposure Exsposutr me Whole Body Skin Ground Plane Inhalation

(Sectors) (meters) ,

Police at Shooting Range SE 700 52.0 5.15E-5 7.63E-5 5.66E-7 4.36E-4

Tour Participants(a) Simulator Bldg. S 310 1.00 6.75E-7 1.00E-6 3.93E-9 5.72E-6

(b) Bio Lab SSE 460 1.50 . -1.01E-6 1.50E-6 8.93E-9 8.58E-6

(c) Overlook E 210 0.25 . 4.70E-7 6.96E-7 1.28E-9 3.98E-6

American Indians NW 200 24.0 1.86E-4 2.76E-4 5.85E-7 1.58E-3

at Burial Grounds NNW 200 24.0 1.30E-4 1.93E-4 3.34E-7 1.11E-3

Ranch Hands driving NW 250 0.25 1.30E-6 1.93E-6 4.30E-9 1.10E-5

cattle around site NNW 350 0.25C;ir¢ : ,.4.98E-7 7.38E-7 1.45E-9 4.22E-6

N 320 s . 0.25,, , ,31E- 7 4.90E-7 7.29E-10 2.80E-6

NNE 450 / 0.25,, I. 1.29E-7 1.92E-7 3.14E-10 1.10E-6

NE 630 - 0:25- _ ', ' 6.58E-8 9.76E-8 1.93E-10 5.58E-7

I

II

'yAl

NOTE: All doses are in mrem. ' : r tW 1- )A --

IPage 47,-.1 , I

I I, 1 " I i,

Enclosure, - PG&E Letter DCL-03-049



RADIATION DOSE DUE TO RELEASE OF RADIOACTIVE GASEOUS EFFLUENTSFOURTH QUARTER, 2002

ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUPS)

External Dose Internal Dose_____________________________________________I ,

,

Noble Gas lodines, Particulates, and Tritium .

Exposure Exposure ExsrSpecific Activity Location Closest Epst .Time D Whole Body Skin Ground Plane Inhalation

(Sectors) (meters) (ours)

Police at Shooting Range SE 700 52.0 5.56E-5 8.80E-5 8.01 E-5 6.81 E-4

Tour Participants(a) Simulator Bldg. S 310, 1.00 7.29E-7 1.15E-6 5.56E-7 8.93E-6

(b) Bio Lab SSE 460 - 1.50 .: -1.09E-6 1.73E-6 1.26E-6 1.34E-5

(c) Overlook E 210 0.25 - 5.08E-7 8.04E-7 1.81 E-7 6.21 E-6

American Indians NW 200 '24.0, 2:01 E-4 3.19E-4 8.27E-5 2.47E-3

at Burial Grounds NNW 200 24.0 ' ,"1.41E-4 2.23E-4 4.72E-5 1.72E-3

Ranch Hands driving NW 250 '0.25 1.41 E-6 2.23E-6 6.08E-7 1.72E-5

cattle around site NNW 350 0.25 - 5.38E-7 8.52E-7 2.05E-7 -65E6

N 320' 0.25 3.58E-7 5.66E-7 1.03E7 4.38E6

NNE 450 0.25 ,1.40E-7 2.21 E-7 4.44E-8 1.71 E-6

NE 630 0.25- 7.12E-8 1.13E-7 2.73E-8 8.71 E-7

''2

NOTE: All doses are in mrem. P 48, I

'' ' - Page 48

, - . - ; �,-- 4 �, I "; �



I AITABLE l ERADIATION DOSE DUE TO RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS

ANNUAL TOTAL, 2002ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUPS)

External Dose |Internal Dose

I - Noble-Gas | lodines, Particulates, andTritium

Exposure Exposure Epsr ie WoeBdSpecific Activity Location Closest Dist (HOS)ie Bdy Skin Ground Plane Inhalation

(Sectors) (m eters)__ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _

Police at Shooting Range SE" 700 208.0 3.33E-4 5.12E-4 2.77E-4 1.73E-3

Tour Participants(a) Simulator Bldg. S 310 4.00 4.37E-6 6.71 E-6 1.92E-6 2.27E-5

. . ,. - ,, 9.. t.

(b) Bio Lab SSE 460 6.00 6.55E-6 1.01E-5 4.37E-6 3.41E-5

(c)Overlook E 210 1.00 3.04E-6 4.67E-6 6.28E-7 1.58E-5

American Indians NW 200 96.0 ' 1.21E-3 1.85E-3 2.86E-4 '6.27E-3

at Burial Grounds NNW 200 96.0 8.44E-4 1.30E-3 1.63E-4 4.39E-3

Ranch Hands driving NW 250 1 8.42E-6 1.29E-5 2.1OE-6 4.38E-5

cattle around site NNW 350 ,1... .. , 3 22E-6 4.95E-6 7.09E-7 1.68E-5

N 320 1,. : ., 2.14E-6 3.29E-6 3.57E-7 1.11E-5

-NNE 450 1. 8.37E-7, 1.29E-6 1.54E-7 4.36E-6

NE 630 -1-.-- . 4.26E-7 6.54E-7 9.46E-8 2.22E-6

1,

I II

I ,I

NOTE: All doses are in mrem.Page 49

I� A b-. . I .. , t �.

II - .


XI. Meteorological Data

,,I

Page 50


Meteorological Data

The hour-by-hour listing of wind speed, wind direction, atmospheric stability andprecipitation is contained on compact disc with this submittal.

Page 51

Attachment 1PG&E Letter DCL-03-049

Attachment I

Radiological Monitoring and Controls Program

(ProcedureCY2, Revision 5)

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CY2

NUCLEAR POWER GENERATION REVISION 5

PROGRAM DIRECTIVE PAGE 1 OF 12

TITLE: Radiological Monitoring and Controls Program

10/08/02EFFECTIVE DATE

CLASSIFICATION: QUALITY RELATEDSPONSORING ORGANIZATION: CHEMISTRY

TABLE OF CONTENTS

SECTION PAGE

PROGRAM OVERVIEW .................. I

APPLICABILITY ................. ..DEFINITIONS ..... :3PROGRAM OBJECTIVES AND REQUIREMENTS ......................................... 3

Program Objectives ......................................... 3

Program Requirements ......................................... 4

Radiological Environmental Monitoring Program ......................................... 4

Radioactive Effluent Control Program ......................................... 4

Offsite Dose Calculation Procedures ......................................... 6

Environmental Radiological Monitoring Procedure ......................................... 7Radwaste Treatment Systems ......................................... 7

Quality Assurance Requirements ......................................... 7

RESPONSIBILITIES ......................................... 8

KEY IMPLEMENTING DOCUMENTS ......................................... 8

Inter-Departmental Administrative Procedures (IDAPs) ......................................... 8

Department-Level Administrative Procedures (DLAPs) ......................................... 8

CLOSELY RELATED PROGRAMS ......................................... 9RECORDS ......................................... 9

APPENDICES ........................................ 10

ATTACHMENTS ........................................ 10

REFERENCES ........................................ 10

SPONSOR ........................................ 10

1. PROGRAM OVERVIEW

It is the policy of nuclear generation that the release of radioactive materials to the environment bein compliance with Federal regulations and be "As Low As Reasonably Achievable" (ALARA).The overall objectives are to protect the health and safety of the public from undue radiationexposure and to minimize the amount of radioactive effluents resulting from the operation of the

plant.

This PD defines the overall policies and general requirements related to the Radiological Monitoring

and Controls Program (RMCP). This includes the Radiological Environmental Monitoring Program

(REMP), and the Radioactive Effluent Controls Program (RECP).

01317005 DOC 01A 1008 1225

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CY2

PROGRAM DIRECTIVE REVISION 5PAGE 2 OF 12


The scope of this PD is focused on the control of releases of radioactive material to the environment,

and minimizing radiological impact on the general public. Radiation protection of plant workers

and visitors within the restricted area of the plant is within the scope of RP1, "Radiation Protection."

Figure I illustrates the hierarchy of procedures associated with this PD.

FIGURE 1

CY2Hierarchy of Procedures

CY2

Radiological Monitoring and Controls Program

IDAPs

Radioactive Effluent Control ProgramEnvironmental Radiological Monitoring Procedure

DLAPs

Department Specific Administrative Controls

01317005.DOC 01A 1008 1225

UNCONTROLLEDTUA P OTO P R R ISSUE FOR USEPACIFIC GAS AND ELECTRIC COMPANY NUMBER CY2



2. APPLICABILITY

This PD is applicable to all persons involved in radioactive effluent control, monitoring, and

management activities. This includes all nuclear generation personnel, personnel matrixed to

nuclear generation from other company organizations, personnel in other company organizations

that are engaged in activities in support of nuclear generation, and contractor personnel that are

working under nuclear generation supervision.

3. DEFINITIONS

3.1 ALARA (acronym for "as low as reasonably achievable") - A term that means making

every reasonable effort to maintain exposures to radiation as far below the dose limits of

10 CFR 20 as is practical consistent with the purpose for which the licensed activity is

undertaken, taking into account the state of technology, the economics of improvements

in relation to state of technology, the economics of improvements in relation to benefits to

the public health and safety, and in relation to utilization of nuclear energy and licensed

materials in the public interest. The specific objectives of achieving ALARA effluents

are based on those described in 10 CFR 50, Appendix I.

3.2 The Radiological Monitoring and Controls Program (RMCP) - Contains the Radioactive

Effluent Controls and Radiological Environmental Monitoring Programs required by

Technical Specifications 5.5.1 and 5.5.4 and descriptions of the information that should

be included in the Annual Radiological Environmental Operating and Annual Radioactive

Effluent Release Reports required by Technical Specifications 5.6.2 and 5.6.3.

3.3 Offsite Dose Calculation Procedure (ODCP) - Contains the methodology and parametersused in the calculation of offsite doses due to radioactive gaseous and liquid effluents and

in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints.

3.4 Environmental Radiological Monitoring Procedure (ERMP) - Contains a description of

sample locations, types of sample locations, methods and frequency of analysis, and

reporting requirements.

4. PROGRAM OBJECTIVES AND REQUIREMENTS

4.1 Program Objectives

The nuclear generation radiological monitoring and controls program is established to

meet the following objectives:

4.1.1 Ensure that systems, methods, and controls are established to meet applicableregulatory requirements and objectives for release of radioactive effluents.

Liquid and gaseous radioactive waste processing systems provide the means

for controlling radioactive releases. It is also important to establishadministrative controls with clear delineation of responsibilities to ensure that

monitoring, measurement, and release activities are properly sequenced,authorized, and controlled.

01317005 DOC OIA 1008 1225

TO FORPACIFIC GAS AND ELECTRIC COMPANY NUMBER CY2PROGRAM DIRECTIVE REVISION 5

PAGE 4 OF 12


4.2 Program Requirements

The basic requirement for the radiological monitoring and controls program shall be tomaintain radioactive releases to the unrestricted areas surrounding the plant inconformance with applicable Federal regulations and ALARA. The following sectionsprovide additional requirements for various elements of the program.

4.2.1 Changes to the RMCP (including ODCP, ERMP and RECP) shall be processedin accordance with the requirements of the plant Technical SpecificationSection 5.5.1.

4.2.2 Radiological Environmental Monitoring Program

a. A Radiological Environmental Monitoring Program (REMP) shall beestablished and maintained to comply with the plant TechnicalSpecification 5.5.1, Radiological Environmental Monitoring Programrequirements. The program shall be provided to monitor the radiationand radionuclides in the environs of the plant, and shall address thefollowing:

1. Monitoring, sampling, analysis, and reporting of radiation andradionuclides in the environment in accordance with themethodology and parameters in the Environmental RadiologicalMonitoring Procedure (ERMP),

2. A Land Use Census to ensure that changes in the use of areas at andbeyond the SITE BOUNDARY are identified and thatmodifications to the monitoring program are made if required bythe results of this census, and

3. Participation in an Interlaboratory Comparison Program to ensurethat independent checks on the precision and accuracy of themeasurements of radioactive materials in the environmental samplematrices are performed as part of the quality assurance program forenvironmental monitoring.

4.2.3 Radioactive Effluent Control Program

a. Monitoring requirements shall be established and maintained for allmajor and potentially significant paths for release of radioactive materialduring normal plant operation, including anticipated operationaloccurrences, to comply with Regulatory Guide 1.21, Revision 1,June 1974, requirements.

01317005.DOC 01A 10081225

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b. Procedures shall be established and maintained to define the methods andrequirements for control of liquid and gaseous radioactive dischargeswithin the limits of the plant Technical Specification Section 5.5.4.These procedures shall address the following:

1. Limitations on the operability of radioactive liquid and gaseousmonitoring instrumentation including surveillance requirements andsetpoint determination in accordance with methodology in the

Offsite Dose Calculation Procedure, (ODCP),

2. Limitations on the concentrations of radioactive material released inliquid effluents to UNRESTRICTED AREAS conforming to10 CFR Part 20, Appendix B, Table 2, Column 2,

3. Monitoring, sampling, and analysis of radioactive liquid andgaseous effluents in accordance with 10 CFR 20.13 02 and with themethodology and parameters in the ODCP,

4. Limitations on the annual and quarterly doses or dose commitmentto a MEMBER OF THE PUBLIC from radioactive materials inliquid effluents released from each unit to UNRESTRICTEDAREAS conforming to Appendix I to 10 CFR Part 50,

5. Determination of cumulative and projected dose contributions fromradioactive effluents for the current calendar quarter and currentcalendar year in accordance with the methodology and parametersin the ODCP at least every 31 days,

6. Limitations on the operability and use of the liquid and gaseouseffluent treatment systems to ensure that the appropriate portions ofthese systems are used to reduce releases of radioactivity when theprojected doses in a 31-day period would exceed 2 percent of theguidelines for the annual dose or dose commitment conforming toAppendix I to 10 CFR Part 50,

7. Limitations on the dose rate resulting from radioactive materialreleased in gaseous effluents from the site to areas at or beyond theSITE BOUNDARY shall be limited to the following:

a) For noble gases: Less than or equal to a dose rate of500 mremlyr to the whole body and less than or equal to adose rate of 3000 mrem/yr to the skin, and

b) For Iodine-I31, for Iodine-133, for tritium, and for allradionuclides in particulate form with half-lives greater than 8days: Less than or equal to a dose rate of 1500 mrem/yr toany organ.

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8. Limitations on the annual and quarterly air doses resulting fromnoble gases released in gaseous effluents from each unit to areasbeyond the SITE BOUNDARY conforming to Appendix I to10 CFR Part 50,

9. Limitations on the annual and quarterly doses to MEMBERS OFTHE PUBLIC from Iodine-13 1, Iodine-133, tritium, and allradionuclides in particulate form with half-lives greater than 8 daysin gaseous effluents released from each unit to areas beyond theSITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,and

10. Limitations on the annual dose or dose commitment to anyMEMBER OF THE PUBLIC due to releases of radioactivity and toradiation from uranium fuel cycle sources conforming to 40 CFRPart 190.

11. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to theRadioactive Effluent Controls Program Surveillance Frequency.

c. Sampling and analysis methods associated with effluent monitoringactivities shall be controlled in accordance with a department leveladministrative procedure that controls material and equipment used foranalysis for the chemistry and radiochemistry programs.

d. Systems that are known pathways for radioactive releases shall beexplicitly addressed. Periodic sampling of systems with the potential ofbecoming radioactively contaminated should also be addressed.

e. An onsite meteorological program shall be established and maintained inaccordance with the requirements of Regulatory Guide 1.23,February 1972, to provide sufficient data for the performance of doseassessments.

f. The collection and processing of technical data required to support theAnnual Radioactive Effluent Release Report and non-routine reports tothe NRC to comply with the plant Technical Specification 5.6.3 shall bedefined as part of this program. The processing of these reports shall beperformed in accordance with XIl, "Regulatory Interface."

4.2.4 Offsite Dose Calculation Procedures

a. Offsite Dose Calculation Procedures (ODCP) shall be established andmaintained to define and control the methods for determining offsitedoses. NRC Regulatory Guide 1.109, Revision 1, October 1977, as wellas its interpretation through NUREG 0133, should be used as guidancefor establishing acceptable methods. These procedures shall address thefollowing:

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1. Methods for determining monitoring instrumentation alarmsetpoints are addressed in accordance with a Department-LevelAdministrative Procedure (DLAP) under CY2.

2. Methods for determining effluent concentrations.

3. Methods for calculating doses to persons in unrestricted areassurrounding the plant from all exposure pathways.

b. Changes to the ODCP shall be processed in accordance with therequirements of Technical Specification 5.5.1.

4.2.5 Environmental Radiological Monitoring Procedure

a. An Environmental Radiological Monitoring Procedure (ERMP) shall beestablished and shall contain a description of sample locations, types ofsample locations, methods and frequency of analysis, and reportingrequirements.

4.2.6 Radwaste Treatment Systems

a. Radwaste Treatment Systems shall be provided to control the processingand release of radioactive materials in gaseous and liquid effluent incompliance with Technical Specification requirements. The design ofthese systems shall be controlled in accordance with CF3, "DesignControl," and the requirements of Regulatory Guide 1.143, October 1979.

b. Approval of changes to the radwaste treatment systems shall beprocessed in accordance with the requirements of CF4, "ModificationControl."

4.2.7 Quality Assurance Requirements

In addition to requirements specified in earlier sections and those requirementsutilizing procedures in the section 6.2 of this PD, the control program shall besubject to the quality assurance requirements specified in CYI,"Chemistry/Radiochemistry."

4.3 Support by the Company Departments

Departments outside of nuclear generation may be cailed upon to support nucleargeneration activities associated with the Radioactive Monitoring Controls Program. Thecontract or agreement between nuclear generation and other departments shall ensure thesupport is performed in accordance with the requirements of this PD.

EXAMPLE: The technical and ecological director may perform radiological laboratoryanalysis for the Radiological Environmental Monitoring Program and prepare and reviewthe Annual Radiological Environmental Operating Report, in accordance with this PD.

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5. RESPONSIBILITIES

5.1 The chief nuclear officer - is responsible for establishing the policy and generalrequirements for the Radiological Monitoring and Controls Program, for providingmanagement support and guidance for the program's implementation, and ensuringcompliance with all regulatory requirements is maintained.

5.2 The plant operations vice president - is responsible for the overall development,implementation, and maintenance of the Radiological Monitoring and Controls Programin accordance with the requirements of this PD.

5.3 The plant services vice president - is responsible for ensuring that support from reportingdepartments is provided for the Radiological Monitoring and Controls Program.

5.4 The station director - is responsible for the direct implementation of the RadiologicalMonitoring and Controls Program with the exception of the design of radwaste treatmentand effluent monitoring systems.

5.5 The engineering director - is responsible for maintaining the design bases for installedplant radwaste treatment and effluent monitoring systems, structures, and components andproviding technical support to the plant for the operation and maintenance of thesesystems.

5.6 The licensing director - is responsible for auditing the Radiological Monitoring andControls Program as outlined in step 3.10 of Appendix 9.1.

5.7 The maintenance director - is responsible for maintaining the radiation monitoringsystems and the hardware and software for the Rad Effluent program.

6. KEY IMPLEMENTING DOCUMENTS

6.1 Inter-Departmental Administrative Procedures (IDAPs)

Inter-Department Administrative Procedures shall be developed to address the followingaspects of the Radiological Monitoring and Controls Program:

6.1.1 An IDAP shall be developed to define the requirements and responsibilitiesassociated with the Radioactive Effluent Control Program.

6.1.2 An IDAP shall be developed to define the requirements and responsibilitiesassociated with the Environmental Radiological Monitoring Procedure.

6.2 Department-Level Administrative Procedures (DLAPs)

Departments responsible for performing activities related to the Radioactive EffluentControl program shall develop DLAPs as appropriate to control program activities.

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7. CLOSELY RELATED PROGRAMS

7.1 Interfaces

This section describes each of the principal interfaces and boundaries between thisProgram Directive and other management processes.

7.1.1 ADIO, "Records"

"Records" provides for the retention of Radiological Monitoring and ControlsProgram records.

7.1.2 CF3, "Design Control"

"Design Control" addresses the implementation of design activities forinstalled radwaste treatment and effluent monitoring systems in accordancewith the requirements of NRC Regulatory Guide 1.143.

7.1.3 CF4, "Modification Control"

"Modification Control" addresses the implementation of modification activitiesfor installed effluent monitoring systems.

7.1.4 CYI, "Chemistry/Radiochemistry"

"Chemistry/Radiochemistry" addresses the methods forchemistry/radiochemistry sampling and analysis of liquid and gaseousradioactive effluents in support of this PD.

7.1.5 OM7, "Problem Resolution"

"Problem Resolution" addresses deficiencies identified during theimplementation of the radioactive effluent control program. OM7 alsoaddresses evaluating nonconformances for reportability in accordance withTechnical Specifications.

7.1.6 TQI, "Personnel Training and Qualification"

"Personnel Training and Qualification" identifies training and qualificationrequirements for personnel.

7.1.7 XI I, "Regulatory Interface"

"Regulatory Interface" addresses the process for required reporting andcommunication with outside agencies.

7.1.8 CY2.ID1, "Radioactive Effluent Controls Program"

"Radioactive Effluent Controls Program" contains the general programrequirements to ensure the requirements of 10 CFR Part 20 and10 CFR Part 50, Appendix I, are met.

8. RECORDS

None

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9. APPENDICES

9.1 Graded Quality Assurance Requirements for Radiological Monitoring and ControlsProgram

10. ATTACHMENTS

None

11. REFERENCES

11.1 Title 10, Code of Federal Regulations,

11.1.1 Part 20, "Standards for Protection Against Radiation"

11.1.2 Part 50, Appendix I,

11.1.3 Part 50, Appendix A, GDC 60, 64,

11.1.4 Part 50.36a, "Technical Specifications on Effluents from Nuclear PowerReactors"

11.2 Title 40, Code of Federal Regulations, "Environmental Radiation Protection Standards forNuclear Power Operations."

11.3 Regulatory Guide 1.109, Revision 1, October 1977, "Calculation of Annual Doses to ManFrom Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliancewith 10 CFR Part 50 Appendix I."

11.4 Regulatory Guide 1.143, October 1979, "Design Guidance for Radioactive WasteManagement Systems, Structures, and Components Installed in Light-Water-CooledNuclear Power Plants."

11.5 Regulatory Guide 1.21, Revision 1, June 1974, "Measuring, Evaluating, and ReportingRadioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid andGaseous Effluent from Light-Water-Cooled Nuclear Power Plants."

11.6 Regulatory Guide 1.23, February 1972, "Onsite Meteorological Programs."

11.7 Regulatory Guide 4.15, Revision 1, February 1979, "Quality Assurance For RadiologicalMonitoring Programs (Normal Operations) - Effluent Streams and the Environment."

11.8 Regulatory Guide 4.1, Revision 1, April 1975, "Programs for Monitoring Radioactivity inthe Environs of Nuclear Power Plants."

11.9 Diablo Canyon Nuclear Power Plant Facility Operating Licenses (Unit 1, Unit 2)

11.10 QA Commitment: FSAR Chapter 17.2.

11.11 QA Commitment: Regulatory Guide 1.33.

12. SPONSOR

John Knemeyer

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APPENDIX 9.1

GRADED QA REQUIREMENTSFOR RADIOLOGICAL MONITORING AND CONTROLS PROGRAM

The basis for these Graded QA requirements is to comply with the regulations of 10 CFR 20, 10 CFR 50,

40 CFR 190, the Technical Specifications and Regulatory Guides 1.21, and 4.15.

1 GRADED ITEMS

Radioactive Effluent monitoring instruments are classified as Category 2 or Category 3 items per

Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess

Plant and Environs Conditions During and Following an Accident." These instruments are'used for

detection and assessment of releases and possibly detection of containment breach with

accomplishment of mitigation of the breach. These items fall under Graded QA requirements.

2. GRADED ACTIVITIES

2.1 Installed radiation monitors required per Technical Specification 3.3.3 for monitoringradioactive effluents during plant operations or accidents shall be calibrated at prescribedintervals.

2.2 Sampling and analysis of liquid and gaseous effluents shall be performed in accordancewith CY1, "Chemistry/Radiochemistry."

2.3 Calculations, computer programs, and procedures for evaluating the dose associated with

radioactive effluents shall be performed in accordance with approved quality relatedprocedures.

3. GRADED REQUIREMENTS

3.1 Effluent releases shall be maintained ALARA and shall be performed in accordance withthe requirements of this Program Directive (CY2) to limit the concentrations, doses and

doserates as specified in DCPP Technical Specification 5.5.4, NRC regulations10 CFR 20, 10 CFR 50 Appendix 1, and EPA regulation 40 CFR 190.

3.2 The Annual Radiological Environmental Operating Report shall be developed inaccordance with Technical Specification 5.6.2.

3.3 The Annual Radioactive Effluent Release Report shall be developed in accordance withTechnical Specification 5.6.3.

3.4 Records that support and document the Radiological Monitoring and Controls Program

shall be controlled in accordance with AD 10, "Records."

3.5 Personnel involved in direct implementation of chemistry/radiochemistry, operations, or

radiation protection activities in support of the Radiological Monitoring and ControlsProgram are qualified in accordance with the requirements of TQ 1, "Personnel Trainingand Qualification." In addition, personnel involved in direct implementation of activitiesin support of the Radiological Environmental Monitoring Program are qualified in

accordance with the requirements of an interdepartmental administrative procedure for

Environmental Radiological Monitoring (ERMP).

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APPENDIX 9.1 (Continued)

3.6 Notifications and reports to and correspondence with regulatory agencies shall be done inaccordance with XII, "Regulatory Interface."

3.7 Written plans, procedures and instructions for implementing Radiological Monitoring andControls Program shall be prepared, processed, and controlled in accordance with ADI,"Administrative Control Program."

3.8 Procurement of quality-related equipment or services shall be in accordance with writtenprocedures. Applicable regulatory requirements, design bases, and any otherrequirements necessary to assure adequate quality shall be included in or invoked byreference in documents for procurement of items or services. Test or acceptancerequirements and documentation to be submitted by the supplier shall be identified in theprocurement documents. Receipt inspection requirements, if required, shall be identifiedin the procurement documents.

3.9 Deficiencies identified during implementation of this program shall be documented andcontrolled in accordance with OM7, "Problem Resolution."

3.10 FSAR Chapter 17.18, "Audits," prescribes the audit frequency for various portions of theRMCP:

3.10.1 The performance of activities required by the quality assurance program for theRadioactive Effluents Control Program shall be audited at least once per24 months unless specified otherwise.

3.10.2 The Radiological Environmental Monitoring Program, implementingprocedures, and program results shall be audited at least once per 24 months.

3.10.3 The Offsite Dose Calculation Procedure and its implementing procedures shallbe audited at least once per 24 months.

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Attachment 2

Radioactive Effluent Controls Program

(Procedure CY2.1D1, Revision 6A)

E* NCNRLE PRCDUEZO TS T E ISSU FOR USEjU*

PACIFIC GAS AND ELECTRIC COMPANYNUCLEAR POWER GENERATIONINTER-DEPARTMENTAL ADMINISTRATIVE PROCEDURE

NUMBER CY2.ID1REVISION 6APAGE 1 OF 39

TITLE: Radioactive Effluent Controls Program


-PROCEDURE CLASSIFICATION: QUALITY RELATEDSPONSORING ORGANIZATION: CHEMISTRY

REVIEW LEVEL: "A"

TABLE OF CONTENTS

SECTION PAGE

SCOPE.....................................................................................................................1.DISCUSSION............................................................................................................. 1RESPONSIBILITIES ............. 2INSTRUCTIONS......................................................................................................... 3RECORDS................................................................................................................. 7APPENDICES............................................................................................................ 7REFERENCES ............. 7SPONSOR................................................................................................................. 7

1. SCOPE

1.1 This procedure contains the general program requirements of the Radioactive EffluentControls Program. This program ensures that the requirements of 10 CFR Part 20 and10 CFR Part 50 Appendix I are met.

2. DISCUSSION

2.1 This procedure provides the general requirements for Radioactive Effluent ControlsProgram in accordance with the Technical Specifications and the implementationGeneric Letter 89-01, "Implementation of Programmatic Controls for RadiologicalEffluent Technical Specifications in the Administrative Controls Section of theTechnical Specifications and the Relocation of Procedural Details of RETS to theOff-Site Dose Calculation Manual or to the Process Control Program."

2.2 The following Technical Specification definitions are applicable: T.S. Section 5.5.1

2.2.1 The Offsite Dose Calculation Manual (ODCM) shall contain themethodology and parameters used in the calculation of offsite doses resultingfrom radioactive gaseous and liquid effluents, in the calculation of gaseousand liquid effluent monitoring alarm and trip setpoints, and in the conduct ofradiological environmental monitoring program; and

2.2.2 The ODCM shall contain the radioactive effluent controls and radiologicalenvironmental monitoring activities, and the description of the informationthat should be included in the Annual Radiological Environmental Operating,and the Radioactive Effluent Release Reports required by TechnicalSpecification 5.6.2 and 5.6.3.

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2.2.3 The Diablo Canyon ODCM is made up of the following procedures:

CAP A-8, "Offsite Dose Calculation Procedure"

CY2.ID1, "Radioactive Effluent Controls Program"

RP1.IDl 1, "Environmental Radiological Monitoring Procedure"

CY2, "Radiological Monitoring and Controls Program"

Changes to each of these procedures shall be processed in accordance withthe requirements of Technical Specification Section 5.5.1.

2.3 The specific methodology and parameters used in the calculation of off-site dosesresulting from radioactive gaseous and liquid effluents and in the calculation of gaseousand liquid effluent monitoring Alarm/Trip Setpoints, is contained in CAP A-8,"Off-Site Dose Calculations Procedure (ODCP)." As such, CAP A-8 is incorporatedin this procedure by reference. Therefore, the requirements for revisions to thisprocedure also apply to CAP A-8.

3. RESPONSIBILITIES

3.1 The chemistry manager is responsible for:

3.1.1 Implementation of the Off-Site Dose Calculation Procedure in a manner thatmeets regulatory requirements and preparing the Annual RadiologicalEffluent Release Report.

3.1.2 Providing direction to the Operations Staff in the processing of radioactivewaste streams.

3.1.3 Ensuring that a comparison of the Annual Radioactive Effluent ReleaseReport and the Annual Radiological Environmental Operating Report isperformed.

3.1.4 Ensuring that dose commitment increases due to the Land Use Census inaccordance with Commitment 6.1.12.1 are determined and communicatedpromptly to radiation protection.

3.2 The radiation protection manager is responsible for:

3.2.1 Ensuring the performance of the annual land use census and that the resultsare provided to Chemistry so that Chemistry can established the doserequirements of Commitment 6.1.12.1.

3.2.2 Ensuring that the results of the annual Land Use Census are provided to TESfor inclusion in the Annual Radiological Environmental Operating Report.

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3.2.3 Ensuring that changes to the Environmental Radiological MonitoringProcedure are provided to Chemistry for inclusion in the AnnualRadiological Effluent Release Report.

3.2.4 Ensuring preparation, review and approval of the Nonroutine RadiologicalEnvironmental Operating Report when required by Commitment 6.1.11.1

3.3 TES is responsible for ensuring that REMP sample results exceeding the criteria ofCommitment 6.1.11.1 are communicated promptly to the chemistry manager and theradiation protection manager at the plant.

4. INSTRUCTIONS

4.1 Administrative Requirements

4.1.1 Appendix 6.1 of this procedure contains the operational requirements of theRadioactive Effluent Controls Program.

4.1.2 The operational requirements are implemented by equipment controlguidelines (reference OPL.DC16), CAP A-8, and XI1ID2, "RegulatoryReporting Requirements and Reporting Process."

a. The Equipment Control Guidelines implement those requirements thatare related to equipment and have specific allowed outage times oroperator actions.

b. CAP A-8 includes the methodology and parameters used in thecalculation of off-site doses resulting from radioactive gaseous andliquid effluents and in the calculation of gaseous and liquid effluentmonitoring alarm/trip setpoints.

c. XIl.1D2 implements the reporting requirements.

4.2 Reporting Requirements

4.2.1 Annual Radioactive Effluent Release Report

a. Report Schedule

1. Annual Radioactive Effluent Release Reports covering theoperation of the unit during the previous calendar year shall besubmitted before May 1 of each year, in accordance with10CR50.36a.

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b. The Annual Radioactive Effluent Release Reports shall include:

1. A summary of the quantities of radioactive liquid and gaseouseffluents and solid waste released from the unit as outlined inRegulatory Guide 1.21, "Measuring, Evaluating, and ReportingRadioactivity in Solid Wastes and Releases of RadioactiveMaterials in Liquid and Gaseous Effluents fromLight-Water-Cooled Nuclear Power Plants," Revision 1, June1974, with data summarized on a quarterly basis following theformat of Appendix B thereof. For solid wastes, the format forTable 3 in Appendix B shall be supplemented with threeadditional categories; class of solid wastes (as defined by10 CFR Part 61), type of container (e.g., LSA, Type A, Type B,Large Quantity) and SOLIDIFICATION agent or absorbent (e.g.,cement, urea formaldehyde);

2. A list and description of unplanned releases as defined in ODCPfrom the site to UNRESTRICTED AREAS of radioactivematerials in gaseous and liquid effluents made during thereporting period;

3. Pursuant to Technical Specification 5.5.1, any changes madeduring the reporting period to the following:

* RP2.DC2, "Process Control Program (PCP)"

* CY2.ID1, "Radioactive Effluent Controls Program (RECP)"

* CY2, "Radiological Monitoring and Controls Program(RMCP)"

* RP1 .ID1 1, "Environmental Radiological Monitoring Program(ERMP)"

* CAP A-8, "Offsite Dose Calculation Procedure (ODCP)"

NOTE: An FSAR update may be used in lieu of the ARERR forcommunicating changes to the NRC, regarding the PCP.

4. A listing of new locations for dose calculations and/orenvironmental monitoring identified by the Land Use Censuspursuant to Appendix 6.1.

5. An explanation as to why the inoperability of liquid or gaseouseffluent monitoring instrumentation was not corrected within thetime specified in Appendix 6. 1; and

6. Description of the events leading to liquid holdup tanks or gasstorage tanks exceeding the limits of ECG 19.1 or ECG 24.3.

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c. A discussion of major changes to the Radwaste Treatment Systems(liquid, gaseous and solid). The discussion of each change shallcontain:

1. A summary of the evaluation that led to the determination that thechange could be made in accordance with 10 CFR 50.59;

2. Sufficient detailed information to totally support the reason forthe change without benefit of additional or supplementalinformation;

3. A detailed description of the equipment, components andprocesses involved and the interfaces with other plant systems;

4. An evaluation of the change which shows the predicted releasesof radioactive materials in liquid and gaseous effluents and/orquantity of solid waste that differ from those previously predictedin the License application and amendments thereto;

5. An evaluation of the change which shows the expected maximum-exposures to a MEMBER OF THE PUBLIC in theUNRESTRICTED AREA and to the general population thatdiffer from those previously estimated in the License applicationand amendments thereto;

6. A comparison of the predicted releases of radioactive materials,in liquid and gaseous effluents and in solid waste, to the actualreleases for the period prior to when the changes are to be make;

7. An estimate of the exposure to plant operating personnel as aresult of the change; and

8. Documentation of the fact that the change was reviewed andfound acceptable by the PSRC.

Otherwise the above information may be submitted as part of theannual FSAR update.

d. In addition, the Annual Radioactive Effluent Release Report shall alsoinclude:

1. An annual summary of hourly meteorological data collected overthe previous year. This annual summary may be either in theform of an hour-by-hour listing on magnetic tape/hard disk orother media of wind speed, wind direction, atmospheric stability,and precipitation (if measured), or in the form of joint frequencydistributions of wind speed, wind direction, and atmosphericstability or the licensee has the option of retaining this summaryof required meteorological data on site in a file that shall beprovided to the NRC upon request;

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2. An assessment of the radiation doses due to the radioactive liquidand gaseous effluents released from the unit or station during theprevious calendar year;

3. An assessment of the radiation doses from radioactive liquid andgaseous effluents to MEMBERS OF THE PUBLIC due to theiractivities inside the SITE BOUNDARY (see FSAR Figure 2.1-2)during the report period;

4. All assumptions used in making these assessments, i.e., specificactivity, exposure time and location. The meteorologicalconditions concurrent with the time of release of radioactivematerials in gaseous effluents, as determined by samplingfrequency and measurement, shall be used for determining thegaseous pathway doses. The assessment of radiation doses shallbe performed in accordance with the methodology and parametersin the OFF-SITE DOSE CALCULATION PROCEDURE(ODCP); and

5. An assessment of radiation doses to the likely most exposedMEMBER OF THE PUBLIC from reactor releases and othernearby uranium fuel cycle sources, including doses from primaryeffluent pathways and direct radiation, for the previous calendaryear to show conformance with 40 CFR Part 190,"Environmental Radiation Protection Standards for NuclearPower Operation." Acceptable methods for calculating the dosecontribution from liquid and gaseous effluents are given inRegulatory Guide 1.109, Rev. 1, October 1977.

e. A single submittal may be made for a multiple unit plant. Thesubmittal should combine those sections that are common to all units atthe plant; however, for units with separate radwaste system, thesubmittal shall specify the releases of radioactive material from-eachunit.

4.3 Revisions to the RECP

4.3.1 The requirements for revision to the RECP also apply to CAP A-8.

4.3.2 The requirements are provided in Technical Specification 5.5.1.

4.4 Major changes to Liquid, Gaseous, and Solid Radwaste Treatment Systems

4.4.1 Major changes to the liquid, gaseous, and solid radwaste treatment systemsshall become effective upon review and acceptance by the PSRC providedthe change could be made in accordance with 10 CFR 50.59.

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5. RECORDS

5.1 Data Sheets and records will be maintained in the Records Management System (RMS)in accordance with CYl.DC1, "Analytical Data Processing Responsibilities."

6. APPENDICES

6.1 Operational Requirements of the Radioactive Effluent Controls Program

6.2 High Alarm Setpoints for Fuel Building and Control Room Ventilation SystemsActuation Instrumentation

7. REFERENCES

7.1 CAP A-8, "Off-site Dose Calculation Procedure (ODCP)."

7.2 CAP A-5, "Liquid Radwaste Discharge Management."

7.3 CAP A-6, "Gaseous Radwaste Discharge Management."

7.4 RPl.ID11, 'Environmental Radiological Monitoring Procedure."

7.5 OP1.DC16, "Control of Plant Equipment Not Required by the TechnicalSpecifications.

7.6 XI1ID2, "Regulatory Reporting Requirements and Reporting Process."

- 7.7 Regulatory Guide 1.21, Revision 1, June 1974.

7.8 Regulatory Guide 1.109, Revision 1, October 1977.

7.9 License Amendment Request 93-04.

7.10 10CFR20.1302

7.11 40CFR190

7.12 10CFR50.36a

7.13 10CFR50 Appendix I

7.14 CY2, "Radiological Monitoring and Controls Program"

7.15 QA Commitment: FSAR Chapter 17.2.

7.16 QA Commitment: Reg Guide 4.15.

8. SPONSOR

John Knemeyer

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APPENDIX 6.1

OPERATIONAL REQUIREMENTS OF THE RADIOACTIVE EFFLUENTCONTROLS PROGRAM

6.1.1 Radioactive Liquid Effluent Monitoring Instrumentation (Also covered by ECG 39.3)

Commitment for Operation

*6.1.1.1 The radioactive liquid effluent monitoring instrumentation channels shown inTable 6.1.1-1 shall be OPERABLE with their Alar m/Trip Setpoints set to ensure thatthe limits of Commitment 6.1.3.1 are not exceeded. The Alarm/Trip Setpoints ofthese channels shall be determined in accordance with the methodology and parametersin the OFF-SITE DOSE CALCULATION PROCEDURE (ODCP).

Applicability: At all times.

Action:

a. With a radioactive liquid effluent monitoring instrumentation channelAlarm/Trip Setpoint less conservative than required by the aboveCommitment, immediately suspend the release of radioactive liquid effluentsmonitored by the affected channel or declare the channel inoperable.

b. With less than the minimum number of radioactive liquid effluent monitoringinstrumentation channels OPERABLE, take the ACTION shown inTable 6.1.1-1. Restore the inoperable instrumentation to OPERABLE statuswithin the time specified in the ACTION, or explain in the next AnnualRadioactive Effluent Release Report why this inoperability was not correctedwithin the time specified.

Surveillance Requirements

6.1.1.2 Each radioactive liquid effluent monitoring instrumentation channel shall bedemonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCECHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at thefrequencies shown in Table 6.1.1-2.

6.1.1.3 At least one saltwater pump shall be determined operating and providing dilution to thedischarge structure at least once per 4 hours whenever dilution is required to meet thelimits of Commitment 6.1.3.1.

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TABLE 6.1. 1-1RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION

MINIMUMCHANNELSOPERABLEINSTRUMENT

1. Radioactivity Monitors Providing Alarm and AutomaticTermination of Release

a. Liquid Radwaste Effluent Line (RM-18)#

b. Steam Generator Blowdown Tank (RM-23)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line (FR-20)#

b. Steam Generator Blowdown Effluent

ACTION

1

I

1

2

1 4

Lines (FR-53)

c. Oily Water Separator EffluentLine (FR-251)#

3. Radioactivity Monitor Not Providing AutomaticTermination of Release

Oily Water Separator Effluent Line (RM-3)#

1

1

4

4

1 3

# This Radioactive Liquid Effluent Monitoring Instrumentation is common to both units.

01192906 DOC 01A 0201.0809

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TABLE 6.1.1-1 (Continued)ACTION STATEMENTS

ACTION 1 With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via this pathway may continue for up to 14days provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Commitment6.1.3.2.

b. At least two technically qualified members of the facility staff independentlyverify the release rate calculations and discharge line valvings.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 2

ACTION 3

ACTION 4

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via this pathway may continue for up to 30days provided grab samples are analyzed for radioactivity (beta or gamma) at a lowerlimit of detection of no more than 10-7 microcuries/ml:

a. At least once per 12 hours when the specific activity of the secondary coolant isgreater than 0.01 microcuries/gram DOSE EQUIVALENT 1-131, or

b. At least once per 24 hours when the specific activity of the secondary coolant isless than or equal to 0.01 microcuries/gram DOSE EQUIVALENT 1-131.

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via this pathway may continue for up to 30days provided that, at least once per 12 hours, grab samples are collected and analyzedfor radioactivity (beta or gamma) at a lower limit of detection of no more than I0-7microcuries/ml or transfer the oily water separator effluent to the Liquid RadwasteTreatment System.

With the number of channels-OPERABLE less than required by the Minimum ChannelsOPERABLE requirement,- effluent releases via this pathway may continue for up to 30days provided the flow rate is estimated at least once per 4 hours during actualreleases. Pump performance curves may be used to estimate flow.

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PAGE 11 OF 39


APPENDIX 6.1. (Continued)

TABLE 6.1.1-2RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE

REQUIREMENTS

ChannelCheck

SourceCheckInstrument

'ChannelCalibratio

n

ChannelFunctiona

ITest

1. Radioactivity Monitors Providing Alarm andAutomatic Termination of Release

a. Liquid Radwaste Effluent Line (RM-18)

b. Steam Generator Blowdown Tank (RM-23)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line (FR-20)

b. Steam Generator Blowdown Effluent Line-(FR-53)

c. Oily Water Separator Effluent Line (FR-251)

3. Radioactivity Monitor Not Providing AutomaticTermination of Release

Oily Water Separator Effluent Line (RM-3)

D

D

D(4)

D(4)

P R(3)

M R(3)

N.A.

N.A.

R

R

R

Q(1)

Q(1)

Q

Q

Q

Q(2)

Daily(4) N.A.

D M R(3)

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TABLE 6.1.1-2TABLE NOTATION

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of thispathway and Control Room alarm annunciation occurs if any of the following conditions exits:

a. Instrument indicates measured levels above the Alarm/Trip Setpoint (isolation andalarm), or

b. Relay control circuit failure (isolation only), orc. Instrument indicates a downscale failure (alarm only), ord. Instrument controls not set in operate mode (alarm only).

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that Control Room alarmannunciation occurs if any of the following conditions exist:

a. Instrument indicates measured levels above the Alarm Setpoint, orb. Circuit failure, orc. Instrument indicates a downscale failure, ord. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the referencestandards certified by the National Bureau of Standards (NBS) or using standards that have beenobtained from suppliers that participate in measurement assurance activities with NBS. Thesestandards shall permit calibrating the system over its intended range of energy and measurementrange.- For subsequent CHANNEL CALIBRATION, sources that have been related to the initialcalibration shall be used.

(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release.CHANNEL CHECK for FR-251 shall be made once per calendar day*, and for FR-20 andFR-53 shall be made at least once per 24 hours on days on which continuous, periodic, or batchreleases are made.

(5) Frequency Notation

Notation FrequencyD At least once per 24 hoursDaily At lease once per calendar day*M At least once per 31 daysQ At least once per 92 daysR At least once per 18 monthsP Completed prior to each releaseN.A. Not Applicable

* The frequency "once per calendar day" could result in two successive channel checks nearly 48 hours

apart over a two day period. This frequency is different from and should not be confused with thefrequency notation "D" (at least once per 24 hours) defined in Technical Specifications.

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6.1.2 Radioactive Gaseous Effluent Monitoring Instrumentation (Also covered by ECG 39.4)


6.1.2.1 The radioactive gaseous effluent monitoring instrumentation channels shown inTable 6.1.2-1 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure thatthe limits of Commitment 6.1.6.1 is not exceeded. The Alarm/Trip Setpoints of thesechannels meeting Commitment 6.1.6.1 shall be determined and adjusted in accordancewith the methodology and parameters in the ODCP.

Applicability: As shown in Table 6.1.2-1.

Action:

a. With a radioactive gaseous effluent monitoring instrumentation channelAlarm/Trip Setpoint less conservative than required by the aboveCommitment, immediately suspend the release of radioactive gaseouseffluents monitored by the affected channel, or declare the channelinoperable.

b. With the number of OPERABLE radioactive gaseous effluent monitoringinstrumentation channels less than the Minimum Channels OPERABLE, takethe ACTION shown in Table 6.1.2-1. Restore the inoperableinstrumentation to OPERABLE status within the time specified in theACTION or explain in the next Annual Radioactive Effluent Release Reportwhy this inoperability was not corrected within the time specified.


6.1.2.2 Each radioactive gaseous effluent monitoring instrumentation channel shall bedemonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCECHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at thefrequencies shown in Table 6.1.2-2.

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PAGE 14 OF 39


APPENDIX -6.1 (Continued)

TABLE 6.1.2-1RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

MinimumChannel

Instrument Operable Applicability Action

1. Gaseous Radwaste System

Noble Gas Activity Monitor - Providing

Alarm and Automatic Termination of Release (RM-22) 1 * 5

2. Plant Vent system

a. Noble Gas Activity Monitor Providing Alarm (RM-14 or 1 * 7RM-14R)

b. Iodine Sampler 1 * 9

c. Particulate Sampler 1 * 9

d. Flow Rate Monitor (FR-12) 1 * 6

e. Iodine Sampler Flow Rate Monitor 1 * 6

3. Containment Purge System

Noble Gas Activity Monitor - Providing Alarm and 2(1) ** 8Automatic Termination of Release (RM-44A or 44B)

(l) 2 channels required in Modes 1, 2, 3 and 4. Only 1 channel required in Mode 6 during CoreAlterations or movement of irradiated fuel within containment.

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APPENDIX 6.1 (Continued) -

TABLE 6.1.2-1 (Continued)TABLE NOTATIONS

*At all times.** MODES 1-4;

containment.

ACTION 5

ACTION 6

ACTION 7

ACTION 8

ACTION 9

also MODE 6 during CORE ALTERATIONS or movement of irradiated fuel within

With the number of channels OPERABLE less than required by the MinimumChannels OPERABLE requirement, the contents of the tank(s) may be released to theenvironment for up to 14 days provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, andb. At least two technically qualified members of the facility staff independently

verify the release rate calculations and discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway.

With the number of channels OPERABLE less than required by the MinimumChannels OPERABLE requirement, effluent releases via this pathway may continue forup to 30 days provided the flow rate is estimated at least once per 4 hours.

With the number of channels OPERABLE less than required by the MinimumChannels OPERABLE requirement, effluent releases via this pathway may continue forup to 30 days provided grab samples are taken at least once per 12 hours and thesesamples are analyzed for radioactivity within 24 hours.

With the number of channels OPERABLE less than required by the MinimumChannels OPERABLE requirement, immediately suspend containment PURGING ofradioactive effluents via this pathway.

With the number of channels OPERABLE less than required by the MinimumChannels OPERABLE requirement, effluent releases via the affected pathway maycontinue for up to 30 days provided samples are continuously collected with auxiliarysampling equipment as required in Commitment Table 6.1.6-1.

NOTE FOR ACTION 9: To respond to the low flow alarm, determine that a simplefix cannot be made and that an auxiliary sampler is needed. Move the sampler in,hook up and verify operation, a maximum of two hours is considered a reasonabletime. Over two hours should be considered as exceeding the time limitation of thecommitment for operation (ECG 39.4).

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INTER-DEPARTMENTAL ADMINISTRATIVE PROCEDURE REVISION 6APAGE 16 OF 39



TABLE 6.1.2-2

-RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCEREQUIREMENTS

ChannelCheck

SourceCheckInstrument

ChannelCalibratio

n

ChannelFunctiona

ITest

Modes forWhich

Surveillance

Is Required

1. Gaseous Radwaste System

Noble Gas Activity Monitor -Providing

Alarm and Automatic Termination ofRelease (RM-22)

2. Plant Vent System

a. Noble Gas Activity MonitorProviding Alarm (RM-14 orRM-14R)

b. Iodine Sampler

c. Particulate Sampler

d. Flow Rate Monitor (FR-12)

e. Iodine Sampler Flow RateMonitor

P P R(3) Q(l) *

D M R(3)

W(4) N.A. N.A.

W(4) -N.A. N.A.

D N.A. R

D N.A. R

Q(2)

N.A.

N.A.

QQ

*

*

*

*

3. Containment Purge System

Noble Gas Activity Monitor -Providing Alarm and AutomaticTermination of Release (RM-44A or_44B)

D - P R(3) Q(1)**

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PAGE 17 OF 39




* At all times.** MODES 1-4; also MODE 6 during CORE ALTERATIONS or movement of irradiated fuel within

containment.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of thispathway and control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm/Trip Setpoint (isolation andalarm), or

b. Instrument indicates a downscale failure (alarm only), orc. Instrument controls not set in operate mode (alarm only).

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarmannunciation occurs if any of the following conditions exist:

a. Instrument indicates measured levels above the Alarm Setpoint, orb. Circuit failure, orc. Instrument indicates a downscale failure, ord. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the referencestandards certified by the National Bureau of Standards (NBS) or using standards that have beenobtained from suppliers that participate in measurement assurance activities with NBS. Thesestandards shall permit calibrating the system over its intended range of energy and measurementrange. For subsequent CHANNEL CALIBRATION, sources that have been related to the initialcalibration shall be used.

(4) The CHANNEL CHECK shall consist of verifying that the iodine cartridge and particulate filterare installed in the sample holders.


Notation FrequencyD At least once per 24 hoursW At least once per 7 daysM At least once per'31 daysQ At least once per 92 daysR At least once per 18 monthsP Completed prior to each releaseN.A. Not Applicable

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6.1.3 Liquid Effluents - CONCENTRATION


6.1.3.1 The concentration of radioactive material released in liquid effluents toUNRESTRICTED AREAS (see TS Figure 5.1-3) shall be limited to theconcentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2for radionuclides other than dissolved or entrained noble gases. Fordissolved or entrained noble gases, the concentration shall be limited to2 x 104 microcurie/ml total activity.


Action:

With the concentration of radioactive material released in liquid effluents toUNRESTRICTED AREAS exceeding the above limits, immediately restorethe concentration to within the above limits.


6.1.3.2 Radioactive liquid wastes shall be sampled and analyzed according to thesampling and analysis program of Table 6.1.3-1.

6.1.3.3 The results of the radioactivity analyses shall be used in accordance with themethodology and parameters in the ODCP to assure that the concentrationsat the point of release are maintained within the limits of Commitment6.1.3.1.

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_ _ _ _ _ _ __

NUMBERREVISIONPAGE

CY2.ID16A19 OF 39



TABLE 6.1.3-1RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM

LIQUID RELEASE SAMPLING MINIMUM TYPE OF LOWER LIMITTYPE FREQUENCY -ANALYSIS ACTIVITY OF DETECTION

FREQUENCY ANALYSIS (LLD) (jiCi/ml)(')

1. Batch Waste P. - PRelease Each Batch Each Batch Principal Gamma 5x107

Tanks (4) Emitters (6)

1-131 lx106

M Dissolved andEntrained Gases 1x10-5

One Batch/M (Gamma emitters)

P .M H-3 -_x10-5

Each Batch Composite(2 ) Gross Alpha 1x10-7

P Q Sr-89, Sr-90 5x10-8

Each Batch Composite(2) Fe-55 1x10-6

2. Continuous D W Principal Gamma -Releases(5) Grab Sample Composite(3) Emitters (6) 5x10-7

I-131 lx106

Steam Generator M M Dissolved andBlowdown Tank Grab Sample Entrained Gases

(Gamma emitters) ,

D M H-3 1x10 5

Grab Sample Composite(3) Gross Alpha 1x10-7

D Q Sr-89, Sr-90 5x1078

Grab Sample, Composite(3) Fe-55 lx104

3. Continuous D WReleases(s Grab Sample Composite(3) Principal Gamma 5x10-'

EmittersOily WaterSeparator Effluent.

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PAGE 20 OF 39




(1) The LLD is defined, for the purposes of these specifications, as the smallest concentration ofradioactive material in a sample that will yield a net count, above system background; that will bedetected with 95 % probability with only 5 % probability of falsely concluding that a blankobservation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

LLD =- 4.66sb

E * V *2.22x106 *Y*exp(-4At)

Where:

LLD = the "a priori" lower limit of detection (microcuries per unit mass or volume),Sb = the standard deviation of the background counting rate or of the counting rate

of a blank sample as appropriate (counts per minute),E = the counting efficiency (counts per disintegration),V = the sample size (units of mass or volume),

- 2.22 x 106 = the number of disintegrations per minute per microcurie,Y = the fractional radiochemical yield, when applicable,X = the radioactive decay constant for the particular radionuclide (sec-'), and

At = the elapsed time between the midpoint of sample collection and the time ofcounting (sec).

Typical values of E, V, Y, and At should be used in the calculation.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representingthe capability of a measurement system and not as an a posterior (after the fact) limit for aparticular measurement.

(2) A composite sample is one in which the quantity of liquid sampled is proportional to the quantityof liquid waste discharged and in which the method of sampling employed results in a specimenwhich is representative of the liquids released.

(3) To be representative of the quantities and concentrations of radioactive materials in liquideffluents, samples shall be composited in proportion to the rate of flow of the effluent stream.Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for thecomposite sample to be representative of the effluent release. -

(4) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling foranalyses, each batch shall be isolated, and then thoroughly mixed, by a method described in theODCP, to assure representative sampling.

01 192906.DOC 01A 0201.0809

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NUMBERREVISIONPAGE

CY2.ID16A21 OF 39

TITLE: Radioactive Effluent Controls Prograin

APPENDIX -6.1 (Continued)


(5) A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from avolume of system that has an input flow during the continuous release.

(6) The principal gamma emitters for which the LLD specification applies include the followingradionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141.Ce-144 shall also be measured but with an LLD of 5x10-6 . This list does not mean that only

* these nuclides are to be considered. Other gamma peaks that are identifiable, together with thoseof the above nuclides, shall also be analyzed and reported in the Annual Radioactive EffluentRelease Report.

(7) Frequency Notation:

NotationDWMQP

FrequencyAt least once per 24 hours.At least once per 7 days.At least once per 31 days.At least once per 92 days.Completed prior to each release.

6.1.4 Liquid Effluents - Dose


6.1.4.1 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactivematerials in liquid effluents released, from each unit, to UNRESTRICTED AREAS(see FSAR Figure 5.1-3) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the wholebody and to less than or equal to 5 mrem to any organ, and

b. During any calendar year to less than or equal to 3 mrem to the whole bodyand to less than or equal to 10 mrem to any organ.


Action:

a. With the calculated dose from the release of radioactive materials in liquideffluents exceeding any of the above limits, prepare and submit to theCommission within 30 days, pursuant to 10 CFR 50.4, a Special Report thatidentifies the cause(s) for exceeding the limit(s) and defines the correctiveactions that have been taken to reduce the releases and the proposedcorrective actions to be taken to assure that subsequent releases will be incompliance with the above limits.

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6.1.4.2 Cumulative dose contributions from liquid effluents for the current calendar quarterand the current calendar year shall be determined in accordance with the methodologyand parameters in the ODCP at least once per 31 days.

6.1.5 Liquid Radwaste Treatment System


6.1.5.1 The Liquid Radwaste Treatment System* shall be OPERABLE and appropriate -portions of the system shall be used to reduce the radioactive materials in liquid wastesprior to their discharge when the projected doses due to the liquid effluent, from eachunit, to UNRESTRICTED AREAS (see FSAR Figure 2.1-2) would exceed 0.06 mremto the whole body or 0.2 Mrem to any organ in a 31-day period.


Action:

a. With any portion of the Liquid Radwaste Treatment System not in operationand with radioactive liquid waste being discharged without treatment and inexcess of the above limits, prepare and submit to the Commission within30 days, pursuant to 10 CFR 50.4, a Special Report which includes thefollowing information:

1. Explanation of why liquid radwaste was being discharged withouttreatment, identification of any inoperable equipment or subsystems,and the reason for the inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLEstatus, and

3. Summary description of action(s) taken to prevent a recurrence.


6.1.5.2 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall beprojected at least once per 31 days, in accordance with the methodology andparameters in the ODCP when Liquid Radwaste Treatment Systems are not being fullyutilized.

6.1.5.3 The installed Liquid Radwaste Treatment System shall be considered OPERABLE bymeeting Commitments 6.1.3.1 and 6.1.4.1.

* The Liquid Radwaste Treatment System is common to both units.

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6.1.6 Gaseous Effluents - Dose Rate


6.1.6.1 The dose rate due to radioactive materials released in gaseous effluents from the site toareas at or beyond the SITE BOUNDARY (see FSAR Figure 2.1-2) shall be limited tothe following:

a. For noble gases: Less than or equal to 500 mrem/yr to the whole body andless than or equal to 3000 mrem/yr to the skin, and

b. For Iodine-131, for Iodine-133, for tritium, and for all radionuclides inparticulate form with half-lives greater than 8 days: Less than or equal to1500 mrem/yr to any organ.


Action:

With the dose rate(s) exceeding the above limits, immediately decrease the release rateto within the above limit(s).

- Surveillance Requirements

6.1.6.2 The dose rate due to noble gases in gaseous effluents shall be determined to be withinthe above limits in accordance with the methodology and procedures of the ODCP.

6.1.6.3 The dose rate due to Iodine-131, Iodine-133, tritium, and all radionuclides inparticulate form with half-lives greater than 8 days in gaseous effluents shall bedetermined to be within the above limits in accordance with the methodology andprocedures of the ODCP by obtaining representative samples and performing analysesin accordance with the sampling and analysis program specified in Table 6.1.6-1.

01 192906.DOC 01A 0201.0809

A_

q UE TO R X S'*S*,UBCOtR~tEDPRCEOU~v2D'7OTPACIFIC GAS AND ELECTRIC COMPANYINTER-DEPARTMENTAL ADMINISTRATIVE PROCEDURE

NUMBERREVISIONPAGE

CY2.ID16A24 OF 39



TABLE 6.1.6-1-RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM

LOWERLIMIT OF

MINIMUM TYPE OF DETECTIONGASEOUS RELEASE SAMPLING ANALYSIS ACTIVITY (LLD)

TYPE FREQUENCY FREQUENCY ANALYSIS (gCi/mil) (1)

1. Waste Gas Decay Tank P P Principal Gamma Ix104

Each Tank Each Tank Emitters( (noble gases)Grab Sample .

2. Containment Purge P P Principal Gamma 1x10 -Each Purge(2) Each Purge(2) Emitters() (noble gases)Grab Sample

I-131, 1-133 - 1x10-9

Principal Gamma 1x10-9Emitters (particulates)-

H-3 1xio4

3. Plant Vent M(2) M(2) Principal Gamma 1x10'Grab Sample Emitters() (noble gases)

w(3)(5 W H-3 1XlO6Grab Sample

4. All Release Types as Continuous(6 ) W(4) I-131 -x10' 2

-listed in 1., 2;, 3.,above, at the-plant vent ,

Charcoal 1-133 1x10-'0

-Sample

Continuous(6) W(4) Principal Gamma 1x10". Particulate Emitters(7)

-Sample __.

Continuous(6) M Gross Alpha 1x10-1

CompositeParticulate

Sample .

Continuous 6)~ Q, Sr-89, Sr-90 1x10-l'CompositeParticulate-Sample

5. Steam Generator M~ s MM Principal Gamma lx1O4Blowdown Tank Vent Emitters() (noble

gases)

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(1) The LLD is defined, for purposes of these specifications, as the smallest concentration ofradioactive material in a sample that will yield a net count, above system background, that will bedetected with 95 % probability with only 5 % probability of falsely concluding that a blankobservation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

LLD =4.66Sb

E * V * 2.22 x 106 * Y * exp (-At)

Where:

LLD = the "a priori" lower limit of detection (microcuries per unit mass orvolume),

Sb = the standard deviation of the background counting rate or of the countingrate of a blank sample as appropriate (counts per minute),

E = the counting efficiency (counts per disintegration),V = the sample size (units of mass or volume),

2.22 x 106 = the number of disintegrations per minute per microcurie,Y = the fractional radiochemical yield, when applicable,

= the radioactive decay constant for the particular radionuclide (sec'), andAt = - the elapsed time between the midpoint of sample collection and the time of

counting (sec).

Typical values of E, V, Y, and At should be used in the calculation.

It should be recognized that the LLD is defined as-an a priori (before the fact) limit representingthe capability of a measurement system and not as an a posterior (after the fact) limit for aparticular measurement.

(2) Sampling and analyses shall also be performed following shutdown, startup, or a THERMALPOWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period.

(3) Tritium grab samples shall be taken a least once per 24 hours when the refueling canal is flooded.

01 192906.DOC 01A 0201.0809

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TABLE 6.1.6-1 (Continued)TABLE NOTATIONS (Continued)

-(4) Samples shall be changed at least once per 7 days and analyses shall be completed within 48hours after changing or after removal from sampler. Sampling shall also be performed at leastonce per 24 hours for at least 7 days following each shutdown, startup or THERMAL POWERchange exceeding 15% of RATED THERMAL POWER within a 1-hour period and analysesshall be completed within 48 hours of changing. When samples collected for 24 hours areanalyzed, the corresponding LLD's may be increased by a factor of 10. This requirement doesnot apply if: (1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the reactorcoolant has not increased more than a factor of 3; and (2) the noble gas monitor shows thateffluent activity has not increased more than a factor of 3.

(5) Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the- spent fuel pool area, whenever 'spent fuel is in the spent fuel pool.

(6) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the timeperiod covered by each dose or dose rate calculation made in accordance with Commitments6.1.6.1, 6.1.7.1, and 6.1.8.1.

(7) The principal gamma emitters for which the LLD specification applies include the followingradionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases andMn-54 , Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144 inIodine and particulate releases. This list does not mean that only these nuclides are to beconsidered. Other gamma peaks that are identifiable, together with those of the above nuclides,shall also be analyzed and reported in the Annual Radioactive Effluent Release Report.

(8) Grab samples shall be taken and analyzed at least once per 31 days whenever there is flowthrough the steam generator blowdown tank. Releases of radioiodines shall be estimated basedon secondary coolant concentration and partitioning factors during releases or shall be measured.


Notation FrequencyW At least once per 7 daysM At least once per 31 daysQ At least once per 92 daysP Completed prior to each release

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6.1.7 Dose - Noble Gases


6.1.7.1 The air dose due to noble gases released in gaseous effluents, from each unit, to areasat or beyond the SITE BOUNDARY (see TS Figure 5.1-3) shall be limited to thefollowing:

a. - During any calendar quarter: Less than or equal to 5 mrad for gammaradiation and less than or equal to 10 mrad for beta radiation, and

b. During any calendar year: Less than or equal to 10 mrad for gammaradiation and less than or equal to 20 mrad for beta radiation.


Action:

a. With the calculated air dose from radioactive noble gases in gaseouseffluents exceeding any of the above limits, prepare and submit to theCommission within 30 days, pursuant to 10 CFR 50.4, a Special Report thatidentifies the cause(s) for exceeding the limit(s), defines the correctiveactions that have been taken to reduce the releases and the proposedcorrective actions to be taken to assure that subsequent releases will be incompliance with the above limits.


6.1.7.2 Cumulative dose contributions for the current calendar quarter and current calendar-year for noble gases shall be determined in accordance with the methodology andparameters in the ODCP at least once per 31 days.

6.1.8 -Dose 'Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form


6.1.8.1 The dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium andall radionuclides in particulate form with half-lives greater than 8 days in gaseouseffluents released, from each unit, to areas at and beyond the SITE BOUNDARY (seeTS Figure 5.1-3) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrem to any organand,

b. During any calendar year: Less than or equal to 15 mrem to any organ.


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Action:

a. With the calculated dose from the release of Iodine-131, Iodine-133, tritium,and radionuclides in particulate form with half-lives greater than 8 days, ingaseous'effluents exceeding any of the above limits, prepare and submit tothe Commission within 30 days, pursuant to 10 CFR 50.4, a Special Reportthat identifies the cause(s) for exceeding the limit(s), defines the correctiveactions that have been taken to reduce the releases and the proposed actionsto be taken to assure that subsequent releases will be in compliance with theabove limits.


6.1.8.2 Cumulative dose contributions for the current calendar quarter and current calendaryear for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form withhalf-lives greater than 8 days shall be determined in accordance with the methodologyand parameters in the ODCP at least once per 31 days.

6.1.9 Gaseous Radwaste Treatment System

, Commitment for Operation

6.1.9.1 The GASEOUS RADWASTE SYSTEM and the VENTILATION EXHAUSTTREATMENT SYSTEM shall be OPERABLE. The appropriate portions of thesesystems shall be used to reduce releases of radioactivity when the projected doses in31 days due to gaseous effluent releases, from each unit, to areas at and beyond theSITE BOUNDARY (see FSAR Figure 2.1-2), would exceed 0.2 mrad to air from'gamma radiation or 0.4 mrad to air from beta radiation or 0.3 mrem to any organ of aMEMBER OF THE PUBLIC.


Action:

a. With radioactive gaseous waste being discharged without treatment and inexcess of the above limits, prepare and submit to the Commission within30 days, pursuant to 10 CFR 50.4, a Special Report that includes thefollowing information:

1. Identification of the inoperable equipment or subsystems and the reasonfor inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLEstatus, and -

3. Summary description of action(s) taken to prevent a recurrence.

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- 6.1.9.2 --Doses due to gaseous releases from each unit to areas at and beyond the SITEBOUNDARY. shall be projected at least once per 31 days, in accordance with themethodology and parameters in the ODCP when Gaseous Radwaste Treatment Systemsare not being fully utilized.

6.1.9.3 The installed VENTILATION EXHAUST TREATMENT SYSTEM and GASEOUSRADWASTE SYSTEM shall be considered OPERABLE by meeting Commitments6.1.6.1 and 6.1.7.1 or 6.1.8.1.

6.1.10 Total Dose


-6.1.10.1 The annual (calendar year) dose or dose commitment to any MEMBER OF THEPUBLIC due to releases of radioactivity and to radiation from uranium fuel cyclesources shall be limited to less than or equal to 25 mrems to the whole body or anyorgan, except the thyroid, which shall be limited to less than or equal to 75 mrems.


Action:

a. With the calculated doses from the release of radioactive materials in liquidor gaseous effluents exceeding twice the limits of Commitments 6.1.4.1 .a,6.1.4.1.b, 6.1.7. .a, 6.1.7.1.b, 6.1.8.1.a, or 6.1.8. .b, calculations shall bemade including direct radiation contributions from the units and from outsidestorage tanks to determine whether the above limits of RegulatoryCommitment 6.1.10.1 have been exceeded. If such is the case, prepare andsubmit to the Commission within 30 days, pursuant to 10 CFR 50.4, aSpecial Report that defines the corrective action to be taken to reducesubsequent releases to prevent recurrence of exceeding the above limits andincludes the schedule for achieving conformance with the above limits. ThisSpecial Report, as defined in 10 CFR 20.2203(a), shall include an analysisthat estimates the radiation exposure (dose) to a MEMBER OF THEPUBLIC from uranium fuel cycle sources, including all effluent pathway anddirect radiation, for the calendar year that includes the release(s) covered bythis report. It shall also describe levels of radiation and concentrations ofradioactive material involved, and the cause of the exposure levels orconcentrations. If the estimated dose(s) exceeds the above limits, and if therelease condition resulting in violation of 40 CFR Part 190 has not alreadybeen corrected, the Special Report shall include a request for a variance inaccordance with the provisions of 40 CFR Part 190. Submittal of the reportis considered a timely request, and a variance is granted until staff action onthe request is complete.

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6.1.10.2 Cumulative dose contributions from liquid and gaseous effluents shall be determined inaccordance with Commitments 6.1.4.2, 6.1.7.2, and 6.1.8.2, and in accordance with

the methodology and parameters in the ODCP.

6.1.10.3 Cumulative dose contributions from direct radiation from the units and from outsidestorage tanks shall be determined in accordance with the methodology and parametersin the ODCP. This requirement is applicable only under conditions set forth inACTION a. of Commitment 6.1.10.1.

6.1.11 Radiological Environmental Monitoring

6.1.11.1 Commitment f6r Operation - The Radiological Environmental Monitoring Programshall be conducted as specified in RP1 .ID1 1, "Environmental Radiological MonitoringProcedure."


Action:

a. With the confirmed level of radioactivity as the result of plant effluents in anenvironmental sampling medium at a specified location exceeding the"Reporting Levels for-Nonroutine Operating Reports" in-RPl.ID11 whenaveraged over any calendar quarter, prepare and submit to the Commission

-within 30 days from the end of the quarter, pursuant to 10 CFR 50.4, aNonroutine Radiological Environmental Operating Report that identifies thecause(s) for exceeding the limit(s) and defines the corrective actions to be -taken to reduce radioactive effluents so that the potential annual dose to aMEMBER OF THE PUBLIC is less than the calendar year limits of --Commitment 6.1.4.1, 6.1.7.1, or 6.1.8.1. A confirmatory reanalysis of theoriginal, a duplicate, or a new sample may be desirable, as appropriate. Theresults of the confirmatory analysis shall be completed at the earliest time

'consistent with the analysis, but in any case within 30 days. When morethan one of the radionuclides from "Reporting Levels for NonroutineOperating Reports" in RP .ID11 are detected in the sampling medium, thisreport shall be submitted if:

-concentration(1) + concentration(2)

reportinglevel(l ) reportinglevel(2)

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When radionuclides other than those in the "Reporting Levels for NonroutineOperating Reports" in RP1.IDI1 are detected and are the result of planteffluents, a Nonroutine Radiological Environmental Operating Report shallbe submitted if the potential annual dose to a MEMBER OF THE PUBLICfrom all radionuclides is equal to or greater than the calendar year limits ofCommitment 6.1.4.1, 6.1.7.1, or 6.1.8.1. This report shall include anevaluation of any release conditions, environmental factors, or other aspectsnecessary to explain the anomalous result.

6.1.12 LAND USE CENSUS

6.1.12.1 A Land Use Census shall be conducted as specified in RP1.ID1 1, "EnvironmentalRadiological Monitoring Procedure."


Action:

a. With a Land Use Census identifying a location(s) that yields a calculateddose or dose commitment greater than the values currently being calculatedin Commitment 6.1.8.2, identify the new location(s) in the next AnnualRadioactive Effluent Release Report.

b. With a Land Use Census identifying-a location(s) that yields a calculateddose or dose commitment (via the same exposure pathway) 20% greater thanat a location from which samples are currently being obtained in accordancewith Commitment 6.1 11. 1, add the new location(s) within 30 days to theRadiological Environmental Monitoring Program given in the ERMP. Thesampling location(s), excluding the control station location, having thelowest calculated dose or dose commitment(s), via the same exposurepathway, may be deleted from this monitoring program after October 31 ofthe year in which this Land Use Census was conducted. Submit in the nextAnnual Radioactive Effluent Release Report documentation for a change inthe ERMP including a revised figure(s) and table(s) for the ERMP reflectingthe new location(s) with information supporting the change in samplinglocations.

Bases

6.1.1 Radioactive Liquid Effluent Monitoring Instrumentation

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable,the releases of radioactive materials in liquid effluents during actual or potential releases of liquideffluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted inaccordance with the methodology and parameters in the ODCP to ensure that the alarm/trip willoccur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of thisinstrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 ofAppendix A to 10 CFR Part 50.

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6.1.2 Radioactive Gaseous Effluent Monitoring Instrumentation

-The radioactive gaseous effluent instrumentation is provided to monitor and control, asapplicable, the releases of radioactive materials in gaseous effluents during actual or potentialreleases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculatedand adjusted in accordance with the methodology and parameters in the ODCP to ensure that thealarm/trip will occur prior to exceeding the limits of NUREG 0133. The OPERABILITY anduse of this instrumentation is consistent with the requirements of General Design Criteria 60, 63,and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitorsused to 'show compliance with the gaseous effluent release requirements of RegulatoryCommitment 6.1.7.1 shall be such that concentrations as low as 1 x 10-51 Ci/ml are measurable.

6.1.3 Liquid Effluents - Concentration

This Regulatory Commitment is provided to ensure that the concentration of radioactive materialsreleased in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentrationlevels specified in 10 CFR Part 20, Appendix B, Table 2, Column 2. This limitation providesadditional assurance that the levels of radioactive materials in bodies of water inUNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectivesof Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of -

-10 CFR 20.1301(e) to the population. The concentration limit for dissolved or entrained noblegases is based upon the assumption that Xe-135 is the controlling radioisotope and its EffluentConcentration Limit (ECL) in air (submersion) was converted to an equivalent concentration inwater using the methods described in International Commission on Radiological Protection(ICRP) Publication 2.

This Regulatory Commitment applies to the release of radioactive materials in liquid effluentsfrom all units at the site.

The required detection capabilities for radioactive materials in liquid waste samples are tabulatedin terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and otherdetection limits can be found in Currie, L.A., "Lower Limit of Detection: Definition andElaboration of a Proposed Position for- Radiological Effluent and Environmental Measurements,"NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revisedannually).

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6.1.4 Liquid Effluents - Dose

This Regulatory Commitment is provided to implement the requirements of Sections II.A, III.Aand IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements theguides set forth in Section II.A of Appendix I. The ACTION statements provide the required=operating flexibility and at the same time implement the guides set forth in Section IV.A of

--Appendix I to assure that the releases of radioactive material in liquid effluents toUNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The dosecalculation methodology and parameters in the ODCP implement the requirements inSection III.A of Appendix I that conformance with the guides of Appendix I be shown bycalculational procedures based on models and data, such that the actual exposure of a MEMBEROF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.The equations specified in the ODCP for calculating the doses due to the actual release rates ofradioactive materials in liquid effluents are consistent with the methodology provided in

.-Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of ReactorEffluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents fromAccidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April

-1977.

'This Regulatory Commitment applies to the release-of radioactive materials in liquid effluentsfrom each unit at the site. For units with shared Radwaste Treatment Systems, the liquid - -effluents from the shared system are to be proportioned among the units sharing that system.

6.1.5 Liquid Radwaste Treatment System

The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will beavailable for use whenever liquid effluents require treatment prior to release to the environment. -

The requirement that the appropriate portions of this system be used when specified providesassurance that the releases of radioactive materials in liquid effluents will be kept "as low as isreasonably achievable." This specification implements the requirements of 10 CFR 50.36a,General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given inSection II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use ofappropriate portions of the Liquid Radwaste Treatment System were specified as a suitablefraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50,for liquid effluents.

This Regulatory Commitment applies to the release of radioactive materials in liquid effluentsfrom each unit at the site. For units with shared Radwaste Treatment Systems, the liquideffluents from the shared system are to be proportioned among the units sharing that system.

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6.1.6 Gaseous Effluents - Dose Rate -

This Regulatory Commitment is provided to ensure that the dose at any time at and beyond theSITE BOUNDARY from gaseous effluents from all units on the site will be within the annualdose limits of NUREG 0133 to UNRESTRICTED AREAS. The annual dose limits are the dosesto be associated with the concentrations of 10 CFR Part 20, Appendix B, Table 2, Column 1.These limits provide reasonable assurance that radioactive material discharged in gaseouseffluents will not result in the exposure of a MEMBER OF THE PUBLIC in anUNRESTRICTED AREA; either within or outside the SITE BOUNDARY,- to averageconcentrations exceeding-the limits to be specified in Appendix B, Table 2 of 10 CFR Part 20(10 CFR Part 20.1302(c)). For MEMBERS OF THE PUBLIC who may at times be within the-SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually besufficiently low to compensate for any increase in the atmospheric diffusion factor above that forthe SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, withthe appropriate occupancy factors, shall be given-in the ODCP. The specified release rate limitsof NUREG 0133 restrict, presently, the corresponding gamma and beta dose rates abovebackground to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less thanorequal to 500 mrems/year to the whole body or to less than or equal to 3000 mrems/year to theskin. These release rate limits also restrict, at all times, the corresponding thyroid dose rateabove background to a child via the inhalation pathway to less than or equal to 1500 mrems/year.

This Regulatory Commitment applies to the release of radioactive materials in gaseous effluentsfrom all units at the site.

The required detection capabilities for radioactive material in gaseous waste samples aretabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, andother detection limits can be found in Currie, L.A., "Lower Limit of Detection: Definition andElaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revisedannually).

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6.1.7 Dose - Noble Gases

This Regulatory Commitment is provided to implement the requirements of Sections II.B, III.Aand IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements theguides set forth in Section II.B of Appendix I. The ACTION statements provide the requiredoperation flexibility and at the same time implement the guides set forth in Section IV.A ofAppendix I to assure that the releases of radioactive material in gaseous effluents toUNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The SurveillanceRequirements implement the requirements in Section III.A of Appendix I that conformance withthe guides of Appendix I be shown by calculational procedures based on models and data suchthat the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways isunlikely to be substantially underestimated. -The dose calculation methodology and parametersestablished in the ODCP for calculating the doses due to the actual release rates of radioactivenoble gases in gaseous effluents are consistent with the methodology provided in RegulatoryGuide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluentsfor the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, -

October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport andDispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"Revision 1, July 1977. The ODCP equations provided for determining the air doses at andbeyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

This Regulatory Commitment applies to the release of radioactive materials in gaseous effluents'-from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseouseffluents from the shared system are proportioned among the units sharing the system.

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6.1.8 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form

This Regulatory Commitment is provided to implement the requirements of Sections II.C, III.A,and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guidesset forth in Section II.C of Appendix I. The ACTION statements provide the required operatingflexibility and at the same time implement the guides set forth in Section IV.A of Appendix I toassure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREASwill be kept "as low as is reasonably achievable." The ODCP calculational methods specified inthe Surveillance Requirements implement the requirements in Section III.A of Appendix I thatconformance with the guides of Appendix I be shown by calculational procedures based onmodels and data, such that the actual exposure of a MEMBER OF THE PUBLIC through

-appropriate pathways is unlikely to be substantially underestimated. The ODCP calculationalmethodology and parameters for calculating the doses due to the actual release rates of the subjectmaterials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculationof Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose ofEvaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 andRegulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion ofGaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,: Revision 1, July1977. These equations also provide for determining the actual doses based upon the historicalaverage atmospheric conditions. The release rate specifications for Iodine-131, Iodine-133,tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependentupon the existing radionuclide pathways to man in the areas at and beyond the SITEBOUNDARY. The pathways that were examined in the development of the calculations were:(1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto greenleafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas wheremilk animals and meat-producing animals graze with consumption of the milk and meat by man,and (4) deposition on the ground with subsequent exposure of man.

This Regulatory Commitment applies to the release of radioactive materials in gaseous effluentsfrom each unit at the site. For units with shared Radwaste Treatment Systems, the gaseouseffluents from the shared system are proportioned among the units sharing that system.

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6.1.9 Gaseous Radwaste Treatment System

The OPERABILITY of the GASEOUS RADWASTE SYSTEM and the VENTILATIONEXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenevergaseous effluents require treatment prior to release to the environment. The requirement that theappropriate portions of these systems be used, when specified, provides reasonable assurance thatthe releases of radioactive materials in gaseous effluents will be kept "as low as is reasonablyachievable." This specification implements the requirements of 10 CFR 50.36a, General DesignCriterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D ofAppendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions ofthe systems were specified as a suitable fraction of the dose design objectives set forth in SectionsII.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

This Regulatory Commitment applies to the release of radioactive materials in gaseous effluentsfrom each unit at the site. For units with shared Radwaste Treatment Systems, the gaseouseffluents from the shared system are proportioned among the units sharing that system.

6.1.10 Total Dose

This Regulatory Commitment is provided to meet the dose limitations of 40 CFR Part 190 thathave been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires thepreparation and submittal of a Special Report whenever the calculated doses due to releases ofradioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the wholebody or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.For sites containing up to four reactors, it is highly unlikely that the resultant dose to aMEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individualreactors remain within twice the dose design objectives of Appendix I, and if direct radiationdoses from the units and from outside storage tanks are kept small. The Special Report willdescribe a course of action that should result in the limitation of the annual dose to a MEMBEROF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report,it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from otheruranium fuel cycle sources is negligible, with the exception that dose contribution from othernuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If thedose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part190, the Special Report with a request for a variance (provided the release conditions resulting inviolation of 40 CFR Part 190 have not already been corrected), in accordance with the provisionsof 40 CFR 190.11 and 10 CFR 20.2203(a), is considered to be a timely request and fulfills therequirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relatesto the limits of 40 CFR Part 190, and does not apply in any way to the other requirements fordose limitation of 10 CFR Part 20, as addressed in Regulatory Commitment 6.1.3.1 and 6.1.6.1.An individual is not considered a MEMBER OF THE PUBLIC during any period in whichhe/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

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APPENDIX 6.2

HIGH ALARM SETPOINTS FOR FUEL BUILDING AND CONTROL ROOM VENTILATIONSYSTEMS ACTUATION INSTRUMENTATION

6.2.1 Fuel Building Ventilation System (FBVS) Instrumentation, RE-58 and RE-59

6.2.1.1 RE-58 Nominal Setpoint < 75 mr/hr

Bases6.2.1.2 Fuel Handling Accident in Fuel Handling Building

a. The basis for the RE-58 high alarm setpoint is to initiate actions to mitigateoffside dose consequences from air borne releases resulting from a fuelhandling accident in the Spent Fuel Pool area. Routing ventilation exhaustfrom the Spent Fuel Pool area through the charcoal filter, thus strippinghalogens (principally iodine isotopes) mitigates offsite dose consequences. Thererouting of the ventilation is accomplished automatically upon receipt of aRE-58 high alarm. Receipt of the high alarm also signals personnel toevacuate the area. PG&E performed a calculation (RA-90-1-0 "High and AlertAlarm Setpoint for RE-58") to base the high alarm setpoint of RE-58 on theairborne radioactivity concentration in the fuel Handling Building for the FSARUpdate Expected Case accident release during a fuel handling accident. TheExpected Case Accident consequence presented in the FSAR Update is a lesssevere, but more probable accident than the FSAR Update Design Basis Casefuel handling accident. This resulted in a more conservative (lower) setpointthan that for the Design Basis Accident Case. This calculation analyzed thedetector sensitivity to the various release nuclides as presented in the FSARUpdate.

b. The high alarm setpoint is set to a value more consistent with the ExpectedCase Accident dose rates which eliminates spurious ESF actuation whilelimiting the offsite consequences due to this accident. A < 75mr/hr setpointequates to a site boundary whole body dose of 1.46mr for the duration of theaccident, which has been-evaluated as being acceptable.

6.2.1.3 Inadvertent Criticality in the Spent Fuel Pool

a. The high density Spent Fuel Pool racks were redesigned to assure that a Keff

equal to or less than 0.95 is maintained with the racks fully loaded with fuel ofthe highest anticipated reactivity in each of two regions, and flooded withunborated water at a temperature corresponding to the highest reactivity.PG&E submitted PG&E Letter No. DCL-85-30, "Re-racking of Spent FuelPools," on September 19, 1985. The results of the criticality analysis fornormal and abnormal operations were evaluated in this report. LAR 85-13(PG&E Letter No. DCL-85-333) was submitted on October 30, 1985, andsummarized the results of the Spent Fuel Pool re-racking report. Postulatedevents that could potentially involve accidental criticality were examined and itwas concluded that the limiting value for criticality (Keff of 0.95) would not beexceeded. Therefore, an inadvertent criticality in the Spent Fuel Pool is notconsidered a credible accident and an evaluation of the effect of raising theactuation setpoint on RE-58 was not required.

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b. Radiation monitor RE-59, which monitors the new fuel storage vault area, alsoprovides indication of inadvertent criticality and changes in dose rate for radiationprotection purposes.

6.2.1.4 RE-59 Nominal Setpoint < 15mr/hr

a. In DCPP FSAR update in Chapter 12, Table 12.1-1, "Plant ZoneClassifications," the Fuel Handling Building areas in which RE-59 is located isclassified as Zone III, "Controlled Assess Requiring Short-Term Occupancy" asindicated in Design Criteria Memorandum DCM-T20.

b. The Zone III design maximum dose rate is < 15mr/hr. If the radiation fluxreaches this value, the high alarm will be actuated and the ventilation mode willchange as indicated above with the RE-58 high alarm actuation.

6.2.2 Control Room Ventilation System (CRVS) Instrumentation, RS-25A and RS-26A

6.2.2.1 RS-25A/RS-26A Nominal Setpoint < 2 mr/hr

Bases6.2.2.2 Per calculations DV-1-23 and DV-2-23 attachment 2:

a. The applicable NRC requirement for the radiation exposure (dose) to personnel inthe Main Control Room is IOCFR20.105 "Permissible Level of Radiation inUnrestricted Areas" (superseded). Section b (1) of 10CFR20.105 limits theradiation dose for unrestricted areas to 2mr/hr and Section b (2) limits the dose to100mr is seven consecutive days. If an operator works in the control room for 48hours per week (12 hours per day and 4 days per week, a normal operator workweek) during a 2mr/hr dose rate, the operator will receive a 96mr dose. This isless than the 100mr/week dose limit of 1OCFR20.105 (superseded). Thus, asetpoint of < 2 mr/hr has been established for the Control Room Air InletRadiation Monitors. If the radiation flux reaches this value at any of thedetectors, a change in the ventilation to Mode 4 will be initiated.

b. 10CFR20.105 has been superseded and any changes to this setpoint will require abasis change to reflect the requirements of 10CFR50 Appendix A Criterion 19.

6.2.3 All of the setpoints are controlled by the setpoint control program CF6.ID1 and require a designchange vehicle (request per CF4.IDI) to change. Actual field setpoints are set more conservativeto account for instrument errors.

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Attachment 3

Off-Site Dose Calculation Procedure

(Procedure CAP A-8 Revision 26)

UNICONTROLLED PROCEDURE - DO NOT USE TO QPERFORMWORK orfISSUEFOR USE

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8

NUCLEAR POWER GENERATION REVISION 26

DIABLO CANYON POWER PLANT PAGE 1 OF 62

CHEMICAL ANALYSIS PROCEDURE UNITS

TITLE: Off-Site Dose Calculations 1 AND 206/04/02

EFFECTIVE DATE

PROCEDURE CLASSIFICATION: QUALITY RELATED

TABLE OF CONTENTS

SECTION PAGE

SCOPE..................................................................................................................... 2

DISCUSSION............................................................................................................. 3RESPONSIBILITIES.................................................................................................... 3PREREQUISITES........................................................................................................ 3

PRECAUTIONS.......................................................................................................... 4

INSTRUCTIONS......................................................................................................... 4Liquid Effluents .......................................................................... 4

Liquid Effluents - Dose Calculation .......................................................................... 410 CFR 20, Appendix B, Table 2, Column 2, Effluent (liquid) Concentration Limit (ECL)Calculation .......................................................................... 6Liquid Effluent Radiation Monitor Set Point Methodology .................................................. 7Dose Projection (for Liquid Effluents) ......................................................................... 13Liquid Limiting Flow Rate Methodology - ECL Based ..................................................... 15Liquid Limiting Flow Rates - LLD Based ..................................................................... 16

Gaseous Effluents .......................................................................... 18Plant Vent Noble Gas Monitor - RE-14 HASP ............................................................ 21Containment Purge - RE-44 HASP .......................................................................... 28

Dose To Critical Receptor Due To Radioiodines, Tritium and Particulates Released in GaseousEffluents .......................................................................... 34

40 CFR 190 Dose Calculations .......................................................................... 42ACCEPTANCE CRITERIA .......................................................................... 47REFERENCES .......................................................................... 47RECORDS................................................................................................................ 47APPENDICES .......................................................................... 48ATTACHMENTS .......................................................................... 48

SPONSOR .......................................................................... 48

Table 6.1 - Typical Liquid Effluent Discharge Pathway Allocation Factors ........................ 8

Table 6.2 - Typical Gaseous Effluent Discharge Pathway Allocation Factors ........ ............. 20Table 6.3 - Expected On-Site Distance and Visitation Times for Members of the Public ...... ... 46

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f* UNOTOLDPOE R ONTUSE ;TO PEROR WOR orISEFRUE2PACIFC GA ANDELECTIC CMPAN

PACIFIC GAS AND ELECTRIC COMPANYDIABLO CANYON POWER PLANT

TITLE: Off-Site Dose Calculations

NUMBERREVISIONPAGE

UNITS

CAP A-8262 OF 62

1 AND 2

1. SCOPE

This procedure describes the methodology for the following:

Effluent RECP or RECP or Tech Spec Surveillance Requirement ImplementsType Tech Spec

Liquids 6.1.1.1 (RECP) Determination of alarm/trip setpoints for RE-18, 10 CFR 20 App. B6.1.3.1 (RECP) RE-23, and RE-3 Table 2, Col. 2

Gases 6.1.2.1 (RECP) Determination of alarmltrip setpoints for RE-22, NUREG 01336.1.6.1 (RECP) RE-14, and RE-14R

Liquids 6.1.3.2 (RECP) Prerelease analyses of effluents 10 CFR 20 App. B6.1.3.3 (RECP) Table 2, Col. 2

Postrelease analysis of effluents

Liquids 6.1.4.2 (RECP) Dose calculations 10 CFR 50 App. I

Liquids 6.1.5.2 (RECP) Dose projections 10 CFR 50 App. I

Gases 6.1.6.2 Dose Rate calculations, Noble Gases, Total Body and NUREG 0133Skin

Gases 6.1.6.3 Dose Rate calculations, lodines, Particulates and NUREG 0133Radionuclides other than Noble Gases, per organ, perage group

Gases 6.1.7.2 (RECP) Noble Gas Air Dose Calculations 10 CFR 50 App. I

Gases 6.1.8.2 (RECP) Iodines, Particulates, and Radionuclides other than 10 CFR 50 App. INoble Gases Organ Dose Calculations per age group

Gases 6.1.9.2 (RECP) Noble Gases, lodines, Particulates, and 10 CFR 50 App. IRadionuclides other than Noble Gases, DoseProjection

Liquids 6.1.10.2 Cumulative Dose from: Liquids, Noble Gases, 40 CFR 190and (RECP) lodines, Particulates, and Radionuclides other thanGases 4.4.2.b.1 Noble Gases per age group, per organ

(RECP)

Direct 6.1.10.3 Direct Radiation Dose Rate and Dose Calculations to 40 CFR 190Radiation (RECP) unrestricted areas due to plant and high radwaste

storage sky-shine

The calculational methodology for doses are based on models and data that make it unlikely tosubstantially underestimate the actual exposure of an individual through any of the appropriatepathways. Tables containing the values for the various parameters used in these expressions arealso included.

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PAGE 3 OF 62

TITLE: Off-Site Dose Calculations UNITS 1 AND 2

2. DISCUSSION

2.1 This procedure is used in support of the Radiological Monitoring and ControlsProgram (RMCP), and Radioactive Effluent Controls Program (RECP), and theportion that deals with routine radioactive liquid and gaseous releases to theunrestricted area. Limits are based on the dose commitment to a member of thegeneral public related to the release of radionuclides through either direct or indirectexposure (e.g., submersion in a cloud of radioactive Noble Gases, radionuclidesdeposited on the ground, direct radiation from radionuclides stored on-site, inhalationof radionuclides or ingestion of radionuclides via a food pathway such as milk, meat,vegetable or fish, etc.).

2.2 The conduct of the Environmental Radiological Monitoring Procedure (ERMP) isfound in RPL.ID1.

2.3 Changes to CAP A-8 shall be processed in accordance with the requirements of DCPPTechnical Specification Section 5.5.1.

3. RESPONSIBILITIES

3.1 The Director, Chemistry is the overseeing authority of responsibility for ensuring thatthe off-site dose calculational procedure (ODCP) meets all RECP and Tech Specrequirements with regards to calculated doses delivered by the plant to the unrestrictedarea surrounding the site.

3.2 The Senior Radiochemistry Engineer assumes the overall responsibility for ensuringthat this procedure's program is followed and implemented where appropriate,especially in regards to RECP or Tech Spec requirements.

3.3 The Radiochemistry Effluents Engineer has the responsibility of correct and timelyimplementation of all the procedure's calculational methodology, where appropriate,for each radioactive effluent released. Furthermore this engineer is responsible for:reviewing the results; cross (spot) checking the calculations; and maintaining anupdated archive of post release calculated doses for annual report purposes.

3.4 The Senior Engineer Tech Maintenance Computer Group assures that any supportingcomputer software is maintained current and compatible with the procedure'scalculational methodology and that the computer hardware is maintained operable at alltimes.

3.5 The Radiochemistry staff engineer provides an oversight of the effluents program'sODCP to: confirm compliance with RECP or Tech Specs; provide technical support;recommend or design improvements to the dose calculational methodology and theeffluent program control; and investigate long-term planning toward effluent relatedactivities and their associated dose calculations.

3.6 Responsibilities as described in CYI, " Chemistry and Radiochemistry," andCYL.DC1, "Analytical Data Processing Responsibilities," apply.

4. PREREQUISITES

Not Applicable

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5. PRECAUTIONS

Not Applicable

6. INSTRUCTIONS

6.1 Liquid Effluents

6.1.1 Liquid Effluents - Dose Calculation

The dose contributions to the total body and each individual organ (bone, liver,thyroid, kidney, lung and GI-LLI) of the maximum exposed individual (adult)due to consumption of saltwater fish and saltwater invertebrate is calculatedfor all radionuclides identified in liquid effluents released to unrestricted areasusing the following expression:

D. = Ft AtZA joC, eA " (1)

Where:

Do = The dose commitment to organ, o, in mrem.

Ft = Near field average dilution factor during the period of therelease. It is defined as:

Ft = Waste Flow

Dilution Flow x Z (2)

Where:

Z = Z is the site specific factor for the mixing effect of thedischarge structure. Specifically, it is the credit taken fordilution which occurs between the discharge structure andthe body of water which contaminates fish or invertebratesin the liquid ingestion pathway. For DCPP Z = 5.

At = The time period for the release in hours.

Ai. = The site specific ingestion dose commitment factor toorgan, o, due to radionuclide, i, in mrem/hr per pCi/ml asdefined by Equation 3.

C, Concentration of radionuclide, i, in the undiluted liquideffluent, in gCi/ml.

; = Decay constant of radionuclide, i.

tm = Time interval between end of sampling and midpoint ofrelease.

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PAGE 5 OF 62


The site specific ingestion dose commitment factor, A,,, is defined as:

A30 =kO(UFBF, +UjBI,)DF (3)

Where:

k. = Units conversion factor of 1.14x10 5 in units ofpCi/pCi x ml/l x yr/hr.

UF = Saltwater fish consumption rate in kilograms of fish peryear. DCPP value for this parameter is 21 kg/yr and istaken from NUREG 0133, Section 4.3.1.

BF, = Saltwater bioaccumulation factor for nuclide, i, in fishflesh in units of pCi/Kg per pCi/l. Values for BFj are takenfrom Table A-I of Reg. Guide 1.109.

U, = Saltwater invertebrate consumption rate in kilograms peryears. DCPP value for this parameter is 5 kg/yr and istaken from NUREG 0133, Section 4.3.1.

BI, = Saltwater bioaccumulation factor for nuclide, i, ininvertebrate flesh in units of pCi/Kg per pCi/l. Values forBI, are taken from Table A-I of Reg. Guide 1.109.

DF, = Adult ingestion dose conversion factor for nuclide, i, inmrem per pCi ingested, from Table E-1 l of RegulatoryGuide 1.109, with the following exceptions: H-3, Br-82,Sb-124 and Sb-125 ingestion dose conversion factors aretaken from NUREG/CR-4013. As-76, Sn-i 13, Sn-i i7mand Sb-122 ingestion dose conversion factors werecalculated by ORNL using ICRP-2 methodology.

The site specific values for A1o are listed in Table 10.1.

Units 1 and 2 share a common liquid radwaste (LRW) treatment system.The effluent doses due to releases discharged via the common LRW areapportioned between the units with 50% credited to Unit 1 and 50% creditedto Unit 2.

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6.1.2 10 CFR 20, Appendix B, Table 2, Column 2, Effluent (liquid) ConcentrationLimit (ECL) Calculation

a. The ECL for the identified mixture of radionuclides in the .jtII. batch ofliquids is calculated as follows:

n

ECL ,= (4)

' E Cy

,= ECLY

Where:

ECL, = The unrestricted area total undiluted ECL for the"j th particular mixture of identifiedradionuclides, in p.Ci/ml.

C,, = The concentration of radionuclide "i," in IiCi/mlfor the "jo" mixture.

ECL,, The ECL in unrestricted area water forradionuclide "i," in general, in gCi/ml (from10 CFR 20, Appendix B, Table 2, Column 2).

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b. The overall ECL for simultaneous discharges is given by Equation 5.

ECL overall = J (5)

Z=1 ECLJ

Where:

ECLoverali = The unrestricted area ECL for the current radionuclidemixture for concurrent "j" discharges (in [iCi/ml).

C, = The total activity concentration for the "ji" individualstream in u.Ci/ml.

ECLj = The total ECL for the *jt" individual mixture (orstream) determined as defined in Equation 4, inpCi/ml.

Oj, = The ratio of an individual discharge "j'" pathwayflowrate to the sum total of all individual undilutedpathway flowrates as defined by:

D Ef'f (6)Zfj

J

Where:

f = Undiluted effluent flowrate for pathway, "j".

6.1.3 Liquid Effluent Radiation Monitor Set Point Methodology

a. Introduction

The DCPP radiological effluent controls program requires that the liquideffluent monitors be operable with their alarm/trip set points set toensure that the effluent concentration limits of 10 CFR 20 are notexceeded.

The alarm/trip set point for the liquid effluent radiation monitors isderived from the concentration limit set forth in Appendix B, Table 2,Column 2 of 10 CFR 20.1001-2404.

The alarm/trip set points are applied at the unrestricted area boundary.The set points take into account appropriate factors for dilution,dispersion, or decay of radioactive materials that may occur between thepoint of discharge and the unrestricted area boundary.

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b. Allocation and Safety Factors

The limits of RECP 6.1.3.1 are site limits which require that the set pointmethodology must ensure simultaneous releases do not exceed the liquideffluent concentration limits of 10 CFR 20 in the unrestricted area. TheDCPP High Alarm Set Point (HASP) methodology makes use of anAllocation Factor (AF) to limit the effluent concentrations fromsimultaneous liquid discharges. The Allocation Factors can be adjustedbased upon operational requirements with the restriction that the sum ofthe Allocation Factors must be less than or equal to 1.

Typical Allocation Factors are shown Table 6.1.

Table 6.1

Typical Liquid Effluent Discharge Pathway Allocation Factors

Discharge Pathway Rad Monitor Allocation Factor (AF)Oily Water Separator RE-3 0.01

Liquid Radwaste System RE-I 8 0.90Steam Generator Blow Down (Unit 1) RE-23 (Ul) 0.04Steam Generator Blow Down (Unit 2) RE-23 (U2) 0.04

Miscellaneous none 0.01

An additional level of conservatism in the HASP methodology isimplemented by the use of a Safety Factor (SF). The Safety Factor isdefined as 0.9 and provides for a High Alarm Set Point at 90% of the10 CFR 20 concentration limits.

c. Tritium Correction Factor

As result of an aggressive liquid radwaste treatment program, the liquideffluents at DCPP typically contain very low levels of gamma emitters.In order to reduce the over all volume of liquid waste discharged, DCPPalso recycles waste water. This recycling results in higher tritiumconcentration in liquid effluents when compared with the low gammaemitter concentrations. As a result, standard HASP methodology resultsin very low set points. In some cases the calculated set points are barelyabove the monitor background.

The liquid HASP methodology used by DCPP uses a Tritium CorrectionFactor (TCF) which assumes a constant, but conservative tritiumconcentration in the liquid effluent. This results in an operationallyreasonable set point while ensuring that the liquid effluent concentrationsreleased to the unrestricted areas do not exceed the limits of 10 CFR 20.

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The Tritium Correction Factor is defined as shown in Equation 7.

TCF=LI - CH3 /ECLH3 (7)

Where:

ECLH3 = effluent concentration limit of tritium

CH3 concentration of tritium in the release mix,pre-dilution (gCi/ml)

F = conservative dilution flow rate (gpm)

f conservative undiluted effluent flow rate (gpm)

The concentration of tritium, CH3, is conservatively estimated.

d. Liquid Effluent Radiation Monitor Set Point Calculations

The High Alarm Set Point (HASP) are calculated to ensure that theliquid effluent concentration limits of 10 CFR 20 are not exceeded. Theset points represent the maximum operational set point. The actual setpoint used by operations will be equal to or less than the actual value asdetermined by the HASP methodology described in this section.

1. Set Point Methodology for RE-3 HASP: Oily Water Separator

Under normal conditions, the Oily Water Separator stream does notcontain any radioactive material. Only in the event that there isprimary to secondary leakage does this become a potential liquideffluent discharge point. In order to insure that no unplanned orunmonitored releases take place by way of the Oily WaterSeparator, RE-3 serves to monitor the discharge even when noactivity has been identified in the effluent. When no significantprimary to secondary leakage is taking place or when no activityhas been identified in the Oily Water Separator, the High AlarmSet Point for RE-3 is calculated as shown in Equation 8.

HASP 3 = 3 x BKGDR 3 (8)

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In the event that primary to secondary leakage results in activitybeing detected in the Oily Water Separator, Equation 9 will be usedto calculate a High Alarm Set Point value. The greater HASPvalue as determnined by Equation 8 or Equation 9 will be used.

HASPE3 = BKGDRE 3 + (AFXSF)x :k C7 [ Flf ]x TCF (9)

Where:

HASPRE-3 = high alarm setpoint for RE-3 (cpm)

BKGDRE-3 = background reading for RE-3 (cpm)

(AF) = allocation factor for the oily water separatoreffluent system from Table 6.1

(SF) = safety factor for RE-3 (0.9)

ky = monitor response factor (cpm/IpCi/ml)

C7 = concentration of gamma emitting isotopes in therelease mix, pre-dilution (pCi/ml)

F = dilution flow rate (gpm)

f = undiluted effluent flow rate (gpm)

C, = concentration of isotope "i," in the release mix,pre-dilution (pCi/ml)

ECLi = effluent concentration limit of isotope "i"

TCF = tritium correction factor as definedby Equation 7.

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TITLE: Off-Site Dose Calculations UNITS I AND 2

2. Set Point Methodology for RE-18 HASP: Liquid RadwasteSystem

The High Alarm Set Point for the RE-i 8 Liquid Radwaste Systemliquid effluent radiation monitor is calculated as shown inEquation 10.

HASPRE 18 = BKGD)gE18 + (AFXSF)x kC e CCf Fx TCF (10)

Where:



(AF) = allocation factor for the liquid radwasteeffluent system from Table 6.1

(SF) = safety factor for RE-18 (0.9)

ky= monitor response factor (cpm/gCi/ml)

CY = concentration of gamma emitting isotopes inthe release mix, pre-dilution (plCi/ml)


f = undiluted effluent flow rate (gpm)

C. = concentration of isotope "i," in the releasemix, pre-dilution (pCi/ml)

ECL, = effluent concentration limit of isotope "I"

TCF = tritium correction factor as defined byEquation 7.

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3. Set Point Methodology for RE-23 HASP: Steam GeneratorBlowdown Tank

The High Alarm Set Point for the RE-23, Steam GeneratorBlowdown Tank liquid effluent radiation monitor, is calculated asshown in Equation I l.

HASPRE-23 = BKGDRE-23 + (AFXSF)X ZkrC{ F/f TCF (11)

-z*H3

Where:



(AF) = allocation factor for the steam generatorblowdown effluent system for each unitfrom Table 6.1

(SF) safety factor for RE-23 (0.9)

Ick = monitor response factor (cpm/I"Ci/ml)

CY = concentration of gamma emitting isotopes inthe release mix, pre-dilution (pCi/ml)


f undiluted effluent flow rate (gpm)

C = concentration of isotope "i," in the releasemix, pre-dilution (ptCi/rnl)

ECLi = effluent concentration limit of isotope "i"

TCF = tritium correction factor as defined byEquation 7.

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6.1.4 Dose Projection (for Liquid Effluents)

The projected dose contributions from each reactor unit due to liquideffluents for the current calendar month, quarter and current calendar yearmust be determined in accordance with the methodology and parameters inthe ODCP at least every 31 days.

The purpose of this is to determine if appropriate treatment of liquidradioactive materials in relation to maintaining releases "as low asreasonably achievable," is necessary.

Projections will be made, at least by the end of each month with attention tothe frequency requirement contained in the radiological effluent controlsprogram.

The projected dose from each reactor unit is given by:

DP PU 2 DP.Com (12)

Where:

DP = Projected Dose.

Dpu = Projected dose attributed to reactor unit, U.

DP Com = Projected dose common to both reactor units.

The 31-day projected dose is calculated by Equation 13.

D 31XD PM + d CM' + d CB (13)

(T+t)

Where:

Df = Monthly Projected Dose

DM = Previous Month's Actual Dose

dCM = Current Month Actual Dose to date

d CB = Projected Dose from Current Batch Release

T = Number of days in the previous month

t = Number of days into the present month

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Projected quarterly doses are determined by Equation 14.

DCQ =dcQ (92 +d- +dAB (14)P A +9 t (T +t)

Where:

DpQ = Projected dose for the current calendarquarter.

dQ = Current quarter to date actual dose.

DA= Previous quarter's actual dose.

dpB = Projected dose as a result of the current

batch release.

T = Number of days in the previous quarter.

t = Number of days into the present quarter.

Projected yearly doses are determined by Equation 15.

Dpcy =Clcy d (366 - t3D A dY A(15)A47d (6- (T+t)

Where:

Dpc = Projected dose for the current calendaryear.

dCy = Current year to date actual dose.

DA= Previous year's actual dose.

dpB = Projected dose as a result of the current

batch release.

T = Number of days in the previous year.

t = Number of days into the present year.

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6.1.5 Liquid Limiting Flow Rate Methodology - ECL Based

The maximum effluent flow rate through monitors RE-3, RE-18, and RE-23as well as for releases from the Condensate Demineralizer Regenerate wastetank or miscellaneous release points is established in order to provide furthercontrol over the effluent releases. The release rate limit is determined by theeffluent concentration and the 10 CFR 20 Effluent Concentration Limits(ECLs) as shown in Equation 16.

f F(AFXSFXTCF)f= c(16)vC,

I*H-3ECL,

Where:

f Maximum operational undiluted liquid radwaste effluentdischarge flow rate (gpm)

F = Expected dilution flow rate (gpm)

AF = allocation factor for the liquid radwaste effluent sourcefrom Table 6.1.

SF = safety factor (0.9).

TCF = tritium correction factor as defined by Equation 7.

CQ = concentration of isotopes "i" in the release mix,pre-dilution (plCi/ml)

ECLi = effluent concentration limit of isotope "i" (piCi/ml)

When the term E C5 = 0 then the Limiting Flow Rate is calculatedi*H-3 ECLI

by:

f = F(AFXSFXTCF) (17)

Where the terms are as previously defined.

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6.1.6 Liquid Limiting Flow Rates - LLD Based

When there is no primary to secondary leakage, the Oily Water Separatorand various miscellaneous release points are assumed to be uncontaminated.Furthermore, in order to establish practical operational flow rate limits forany sources when they are considered uncontaminated, Equation 18 is used.While no activity may be present, Equation 18 assumes a concentration equalto the Lower Limit of Detection for the nuclides listed in CY2.ID1,Appendix 6.1, Table 6.1.3-1.

F(A4FXSF) (18)4.3

Where:

f = Maximum operational undiluted liquid radwaste effluentdischarge flow rate (gpm)

F = Expected dilution flow rate (gpm)

AF = allocation factor for the liquid radwaste effluent sourcefrom Table 6.1.

SF = safety factor (0.9).

4.3 Total ECL fraction as given by:

LLD,

ECL,

Where:

LLD, = Lower limit of detection for isotope "i" from CY2.1DI,Appendix 6.1, Table 6.1.3-1. (ttCi/ml)

ECL, = effluent concentration limit of isotope "i" (pCi/mI)

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6.1.7 Unplanned Liquid Releases (Abnormal Releases)

An unplanned release is an unexpected and potentially unmonitored releaseto the environment due to operational error or equipment malfunctions.

a. Unmonitored unplanned releases shall have a report written by theRadiochemistry Effluents Engineer describing the event with acalculation, if possible, of the percent of Tech Spec release rate limit.This will then be forwarded to PSRC for review. Describe theseunplanned releases in the Annual Radioactive Effluent Release Report.

b. Monitored unplanned releases which exceed 1 % of the RECP releaserate limit will also have a report written describing the event and mustbe forwarded to the PSRC for review. Describe these unplannedreleases in the Annual Radioactive Effluent Release Report.

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6.2 Gaseous Effluents

The only significant path for gaseous radioactive releases to the environment duringnormal operations is via the plant vent. This source is used for calculating dose ratesand real-time doses to the unrestricted area due to noble gases, vaporous radioiodinesand airborne radio-particulates. The plant vent also has redundant monitoring for thesetypes of gaseous releases.

Other paths such as the steam generator blowdown tank vent, the chemistry lab fumehood, the main condenser Nash vacuum pump discharge, hot machine shop vent, etc.,are considered miscellaneous release sources. These miscellaneous release sources arenot continuously monitored but can have dose rates and dose calculated for their pathto the unrestricted area. .

6.2.1 Meteorological Methodology

The equations for determining gaseous effluent concentration limits, highalarm setpoints, dose rates, and critical receptor doses make use of thehistorical average atmospheric conditions in accordance with methodologiesof Regulatory Guides 1.109 and 1.111 and NUREGs 0133 and 0472. Thehistorical average dispersion (X/Q) and deposition (D/Q) values are derivedfrom the methodology of Regulatory Guide 1.111 as implemented byNUREG 2919 (computer code XOQDOQ). The DCPP dispersion anddeposition values are based on the latest five years of meteorological dataand are updated when the value of X/Q or D/Q changes by more than tenpercent. The present values are listed in Table 10.2.

Long-term releases are characterized as those that are generally continuousand stable in release rate, such as normal ventilation systems effluents.Doses due to long-term releases are modeled using historical annual averagedispersion and deposition values in accordance with the guidance ofRegulatory Guide 1.109, Regulatory Guide 1.111, NUREG 0133 andNUREG 0472.

Short-term releases are defined as those which occur for a total of 500 hoursor less in a calendar year but not more than 150 hours in any quarter. Inaccordance with NUREG 0133 and based upon an operational history thathas demonstrated short term gaseous releases can be characterized as randomin both time of day and duration, historical average atmospheric dispersionand deposition values are used to model doses due to short-term releases.

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6.2.2 Gas Effluent Concentration Limits

a. Philosophy of Concentration Limits

The radiological effluent controls restrict at all times the dose rate due toradioactive materials released in gaseous effluents from the site to areasat or beyond the site boundary for noble gases to less than or equal to500 mrem/yr to the total body and 3000 mrem/yr to the skin. Foriodine-131, iodine-133, tritium and for all radionuclides in particulateform with half-lives greater than 8 days, the dose rate is limited to lessthan or equal to 1500 mrem/yr to any organ.

These dose rate limits act to restrict at all times the instantaneousconcentrations of radionuclides in gaseous effluents at the site boundary.

1. Allocation and Safety Factors

The limits set forth by RECP 6.1.6.1 are site limits which requirethat the set point methodology must ensure simultaneous releasesdo not exceed the off-site dose rate limits set forth byRECP 6.1.6.1(a) and 6.1.6.1(b). The DCPP High Alarm Set Pointmethodology makes use of an Allocation Factor (AF) to limit thenoble gas effluent dose rate from simultaneous atmosphericreleases.

The Allocation Factors can be adjusted based upon operationalrequirements with the following restrictions:

The sum of the Allocation Factors for RE-14 (plant ventnoble gas monitor), the SGBD tank vents, and miscellaneousrelease points from both units must be less than or equal to 1.

The Allocation Factors for RE-22 (Waste Gas Decay Tanks)and RE-44 (Containment Purge) can also be adjusted basedupon operational requirements with restriction that the sum ofthe Allocation Factors for RE-22 and RE-44 must be less thanor equal to the Allocation Factor for RE-14.

The Allocation Factors for RE-24 (Plant Vent IodineMonitor) and RE-28 (Plant Vent Particulate Monitor) are setequal to the Allocation Factor for RE-14.

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Typical Allocation Factors are shown in Table 6.2.

Table 6.2

Typical Gaseous Effluent Discharge Pathway Allocation Factors

Discharge Pathway Rad Monitor Allocation Factor (AF)Plant Vent - NG Monitor RE-14 0.48Plant Vent Iodine Monitor RE-24 0.48Plant Vent Part Monitor RE-28 0.48,Waste Gas Decay Tank RE-22 0.10Containment Purge RE-44 0.38SGBD Tank Vent 0.01Miscellaneous 0.01

An additional level of conservatism in the HASP methodology isimplemented by the use of a Safety Factor (SF). The Safety Factoris defined as 0.9 and provides for a High Alarm Set Point at 90% ofthe dose rate limits.

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b. Gaseous Effluent Radiation Monitor Set Points

1. PLANT VENT NOBLE GAS MONITOR - RE-14 HASP

The Plant Vent effluent stream is monitored by rad monitor RE-14.RE-14 provides alarm function only.

The High Alarm Set Point methodology for RE-14 is given byEquation 19, which is based upon the assumption that the totalbody dose rate limit is most limiting.

4MaxC (NG)= 500A(G),FxSFx (19)472xFp, x6.Q)fx x 294

Where:

I4MaxCT(NG) = the maximum allowable total concentration of noble gas inthe plant vent, in pCi/cc

AF = the allocation factor for the plant vent for one unit fromTable 6.2

SF = a safety factor to ensure that dose rate limits of theradiological effluent controls are not exceeded (0.9)

500 = (mremlyr) the site Total Body dose rate limit forinstantaneous releases

472 = the conversion constant to cc/sec from cfm

F~pv = total flow rate in the plant vent, in cfm (maximum plantvent flow rate is 263,000 cfm)

VX-QLt = the maximum historical site boundary dispersion factor,

based on 5 year averages derived from the meteorologicaldata base, from Table 10.2.

294 = the whole body dose factor (mrem/yr/jCi/m3) for Xe-133as presented in Table 10.3, (for the plant vent HASP, therelease is assumed to be all Xe-133)

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2. PLANT VENT NOBLE GAS MONITOR - RE-14 SCALING

In order to correlate the readings of RE-14 to noble gasconcentration during periods between samplings, the concentrationis scaled according to Equation 20.

CT = CPM xCS (20)CPMS

Where:

CPMT = RE-14 time weighted arithmetic mean (cpm).

CPMs = RE-14 gross count rate at the time of sampling (cpm).

Cs = Concentration of noble gas corresponding to CPMs,based upon noble gas grab sample (u±Ci/cc).

CT = Scaled concentration of noble gas (gCi/cc).

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3. PLANT VENT IODINE MONITOR - RE-24

The Plant Vent Iodine concentration is monitored by rad monitorRE-24. RE-24 provides alarm function only. The alarm setpointmethodology is based upon the assumption that RE-24 respondsonly to 1-131. The methodology also presumes a release mixturebased upon the RCS source term.

The High Alarm Set Point methodology of RE-24 is given byEquation 21.

24Wfac (Iodine) = SF x AF x f -131 1500 (21)472 x F,, x (jz/p).P,'

Where:

24 "CT(Iodine) = the maximum allowable concentration ofI-131 in the plant vent

AF = The allocation factor for the plant vent forone unit from Table 6.2

SF = A safety factor to insure that the dose ratelimits of the radiological effluent controlsare not exceeded (0.9).

fi11 = fraction of the total non-noble gasconcentration that is due to I-131. Definedas:

= CI_131 (22)

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1500 = (mrem/yr) the site organ dose rate limit forIodine-131, for Iodine-133, for tritium, andfor all radioactive materials in particulateform with half-lives greater than 8 days.


Fpv = total flowrate in the plant vent, in cfm(maximum plant vent flowrate is263,000 cfm)

(il-QL = maximum historical site boundarydispersion factor, based on 5 year averagesderived from the meteorological database,from Table 10.2.

R' = Inhalation dose factor for nuclide "i"(mrem/yr/jiCi/m3) for child age group forworst case organ, from Table 10.4. Dosefactors are based upon NUREG 0133methodology. Inhalation dose conversionfactors are taken from Reg. Guide 1.109,Rev 1, Table E-9, with the followingexceptions: H-3, Sb-124 and Sb-125inhalation dose conversion factors takenfrom NUREG/CR-4013.

f, = fraction of total non-noble gasconcentration (excluding tritium) that is dueto nuclide, i, and defined as:

_ C, (23)EC,

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4. PLANT VENT PARTICULATE MONITOR - RE-28

The Plant Vent Particulate concentration is monitored by radmonitor RE-28. The alarm setpoint methodology is based upon theassumption of a 5% cross talk from the iodine channel. This is dueto the retention of a small portion of iodine on the particulate filter.A release mixture based upon the RCS source term is alsoassumed.

The High Alarm Set Point methodology for RE-28 is given byEquation 24.

28MrCT (Particulates)= AFX SFx (0-05 x >fiodes + ParaaculatesI)lodines Particulates

1500 (24)

X 472x Fv x (IxQ)fmZPf

Where:

" MCT(Particulate) = Maximum allowable particulate concentration inthe plant vent

AF = The allocation factor for the plant vent for oneunit from Table 6.2

SF = A safety factor to insure that the dose rate limitsof the radiological effluent controls are notexceeded (0.9).

0.05 = Fraction of total iodine activity retained onparticulate filter.

flodines = Fraction of the total non-noble gas concentrationthat is due to iodines.

fparuculates = Fraction of the total non-noble gas concentrationthat is due to particulates.

1500 = (mremlyr) the site organ dose rate limit forIodine-131, for Iodine-133, for tritium, and forall radioactive materials in particulate form withhalf-lives greater than 8 days.

472 = Conversion constant to cc/sec from cfm

Fpv = Total flowrate in the plant vent, in cfm(maximum plant vent flowrate is 263,000 cfm)

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VI-Q)- = Maximum historical site boundary dispersionfactor, based on 5 year averages derived from themeteorological database, from Table 10.2

pAw = Inhalation dose factor for nuclide "i"(mrem/yr/gCi/m3) for child age group for worstcase organ, from Table 10.4. Dose factors arebased upon NUREG 0133 methodology.Inhalation dose conversion factors are taken fromReg. Guide 1.109, Rev 1, Table E-9, with thefollowing exceptions: H-3, Sb-124 and Sb-125inhalation dose conversion factors taken fromNUREG/CR-4013.

= Fraction of total non-noble gas concentration(excluding tritium) that is due to nuclide, i, asdefined by Equation 23.

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5. WASTE GAS DECAY TANK MONITOR - RE-22 HASP

Effluent releases from the Waste Gas Decay Tank are monitored byrad monitor RE-22. RE-22 provides alarm and automatic releasetermination functions.

The High Alarm Set Point methodology for RE-22 is given byEquation 25, which is based upon the assumption that the skin doserate limit is most limiting.

22MfC (NG)= AFx SFx 3000

472xF( -ax xl.34x 103(25)

Where:

2aCT(NG) = the maximum allowable total concentration ofnoble gas in the gas decay tank exhaust line, inpCi/cc

AF = the allocation factor for the waste gas decay tankfor one unit from Table 6.2

SF = safety factor to ensure that dose rate limits of theradiological effluent controls are not exceeded(0.9)

3000 = (mrem/yr) the site Skin dose rate limit forinstantaneous releases

472 = the conversion constant to cc/sec from cfmFgdt = maximum flow rate for the gas decay tank system

(31 cfm)

/QlOuaX = the maximum historical site boundary dispersionfactor, based on 5 year averages derived from themeteorological data base from Table 10.2.

1.34x103 = the skin dose factor for Kr-85 (mrem/yr/ pCi/m3 ,from Table 10.3.

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6. CONTAINMENT PURGE - RE-44 HASP

The Containment Purge is monitored by rad monitor RE-44. RE-44provides alarm and automatic release termination functions.

The HASP for RE-44 must limit the noble gas dose rate for skinand total body exposure. In order to address this, two set points arecalculated. One set point is calculated based upon limiting the totalbody dose rate and the other limits the skin dose rate. The morelimiting set point is used. The High Alarm Set Point methodologyfor RE-44 is given by Equations 26 and 27.

a) Limiting Concentration Based on Total Body Dose

44AIBMarCT (NG)= AF x SF x _00 (26)472xF,, x(x/)=xK

Where:

44I'BMXCT(NG) = the maximum allowable total concentration ofnoble gas in the containment purge exhaust line,in pCi/cc

AF = the allocation factor for the containment purgefor one unit from Table 6.2

SF = safety factor to ensure that dose rate limits ofradiological effluent controls are not exceeded(0.9)

500 = (mrem/yr) the site Total Body dose rate limit forinstantaneous releases


Fc = maximum flow rate in the containment purgesystem (maximum containment purge flowrate is55,000 cfm) -

(/QL= the maximum historical site boundary dispersionfactor, based on 5 year averages derived fromthe meteorological data base averages, fromTable 10.2.

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K = whole body dose factor (mrem/yr/jiCi/m3 ) forisotope "i." Dose factors are presented inTable 10.3.

f = the fraction of the concentration of the individualnoble gas radionuclide, "i," in the total mix ofnoble gas effluents in the containment purgeline.

b) Limiting Concentration Based on Skin Dose

Skin dose should be calculated whenever an RCS samplecontains a significant fraction of Kr-85.

The concentration limit calculated by this method shouldthen be compared to the concentration limit calculated fortotal body dose. The smaller concentration should beselected as the limiting concentration for the HASP.

44AIBVaxC (NG)= AF x SF x _ (27)472xF,, xF1Q),,.axZ(L, +l.lM,)f,

Where:

44AIBMaxCT(NG) = the maximum allowable total concentration ofnoble gas in the plant vent in ipCi/cc

AF = the allocation factor for the containment purgefor one unit from Table 6.2

SF = safety factor to ensure that dose rate limits of theradiological effluent controls are not exceeded(0.9)

3000 = (mrem/yr) the site skin dose rate limit forinstantaneous releases.


Fct= total flow rate in the containment purge system,in cfm (maximum containment purge flow rate is55,000 cfm)

/QLj = the maximum historical site boundary dispersionfactor, based on 5 year averages derived fromthe meteorological data base averages, fromTable 10.2.

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Li = the skin dose factor (mrem/yr per iCi/m3 ) forisotope "i." Dose factors are presented inTable 10.3.

1.1 = Conversion factor mrem/mrad. Converts airdose to skin dose.

M = the gamma air dose factor (mrad/yr per uICi/m 3)for isotope 'i." Dose factors are presented inTable 10.3.

f, = the fraction of the concentration of the individualnoble gas radionuclide, "i," in the total mix ofnoble gas effluents in the containment purge line.

c. Mode 6 Particulate activity.

The HASP calculation specified in this section based upon Noble Gaseffluent limitations conservatively bounds the Tech. Spec. requirementfor particulate activity in Mode 6. The FSAR expected case accidentfor Mode 6 is a containment fuel handling accident which does notinclude a particulate release. Therefore, the HASP for RM-44 in thissection conservatively satisfies the Tech. Spec. (Ref.: 8.15)

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6.2.3 Gaseous Dose Rate Calculation Methodology

a. Total Body Noble Gas Dose Rate Methodology

The dose rate to the total body due to immersion in a cloud of noblegases is given by:

V/Q)L K,0, < 500 mrem/year (28)

Where:

.= The release rate of radionuclide i in units of

pCi/sec.

All other terms are as previously defined.

b. Skin Dose Rate Methodology

The dose rate to the skin due to immersion in a cloud of noble gases isgiven by:

(/Q)Af=Z(LI +1 .lM,)Q <3000 mrem/year (29)

Where the terms are as previously defined.

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C. Radioiodine, Tritium and Particulate Dose Rate Methodology

The dose rate to organ, o, due to radioiodines, tritium and particulatesreleased in gaseous effluents is given by:

(x /QL)a P, -Q<<1500 mrem/year (30)

Where:

P,0 K'(BR)DFAIo

and

K' = 106 pCi/pCi

BR = The breathing rate of the a child age group inm3/yr. The default value of 3700 m3/yr is takenfrom Table E-5 of Reg. Guide 1.109.

DFA, = The inhalation dose factor for organ o, for the childage group for radionuclide, i, from Table E-9 ofReg. Guide 1.109 in mrem/pCi, with the followingexceptions: H-3, Sb-124 and Sb-125 inhalationdose conversion factors taken fromNUREG/CR4013.


Values for P,0 are listed in Table 10.6.

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6.2.4 Noble Gas Air Dose Calculation Methodology

a. Gamma Air Dose

The gamma air dose due to immersion in a cloud of noble gases is givenby:

Day =3.17x10-8/Q)AQ-L mIQI (31)

Where:

Day = Gamma air dose in mrad.

3.17x10-8 = Conversion constant yr/sec.

*M. = Gamma air dose factor for nuclide i, in mrad/yrper piCi/m3 . Values are listed in Table 10.3.

Q. = Total release of noble gas radionuclide, i, in gCi.


b. Beta Air Dose

The beta air dose due to immersion in a cloud of noble gases is given by:

Day =3.17x100-(i/Q)-Af N.Q. (32)

Where:

Dap = Beta air dose in mrad.

N. = Beta air dose factor for nuclide i, in mrad/yr perptCi/m 3. Values are listed in Table 10.3.


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6.2.5 Dose To Critical Receptor Due To Radioiodines, Tritium and ParticulatesReleased in Gaseous Effluents

a. Calculation Methodology

The dose to an individual (critical receptor) due to radioiodines, tritiumand particulates released in gaseous effluents with half-lives greater than8 days is determined based upon the methodology described inNUREG 0133. This methodology makes use of the maximum individualconcept described in Regulatory Guide 1.109. The maximum individualis characterized as maximum with regard to food consumption,occupancy, and other usage parameters. This concept therefore modelsthose individuals within the local population with habits representingreasonable deviations from the average. In all physiological andmetabolic respects, the maximum individual is assumed to have thosecharacteristics that represent the average for the age group of interest.

The concept of critical receptor is introduced as a further refinement ofthe maximum individual. The critical receptor is defined as thatindividual that receives the largest dose based upon the combination ofdose pathways that have been shown to actually exist. The criticalreceptor concept is applied at that location where the combination ofdispersion (X/Q), deposition (D/Q), existing pathways, occupancy time,receptor age group, and effluent source term indicates the maximumpotential exposure. The inhalation and ground plane exposure pathwaysare considered to exist at all locations. The grass-cow-milk,grass-cow-meat, and vegetation pathways are considered based on theiractual existence in the vicinity of the plant.

The dose pathways that have been shown to actually exist at DCPP arethe ground plane, inhalation and the vegetation pathways. These dosepathways are reviewed yearly and updated based upon the annual landuse census survey in order to insure that actual exposure to an individualwill not be substantially underestimated.

The locations of the pathways and descriptions are listed in Table 10.2.

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b. Dose Calculation

The dose contributions to the total body and each individual organ(bone, liver, thyroid, kidney, lung and GI-LLI) of the maximum exposedindividual (Critical Receptor) due to radioactive gaseous effluentreleases is calculated for all radionuclides identified in gaseous effluentsreleased to unrestricted areas using the following expression:

Dapo =3.l7xl0 8WcRZRaipOQA (33)

Where:

Dapo = Dose to the critical receptor for age group a,pathway p, and organ o, in mrem.

WCR = Critical receptor z/Q for immersion, inhalation and

all tritium pathways (seconds/m 3) from Table 10.2.

= Critical receptorD/Q for ground plane and all

ingestion pathways (1/m 2) from Table 10.2.

Ralpo = Site specific dose factor for age group a, radionuclide i,pathway p, and organ j (mrem/yr per gCi/m3 forinhalation and tritium pathways - mrem/yr per gCi/(secm2) for ground plane and ingestion pathways). Thesedose factors are listed in Table 10.6.

The site specific dose factors are calculated based uponNUREG 0133 methodology. All dose conversionfactors are taken from Reg. Guide 1.109, Rev 1, TablesE6-E14, with the following exceptions: H-3, Sb-124and Sb-125 dose conversion factors taken fromNUREG/CR-40 13.

Q = The total release of radionuclide i, in units of ,tCi.

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6.2.6 Noble Gas Gaseous Radioactive Waste (GRW) Batch Release PercentRelease Rate Limits (PRRLs) and Expected Reading (ER)

The Percent Release Rate Limit (PRRL) for noble gas releases for eachunit is calculated based upon the 500 mrem/yr whole body dose rate limit,and is given by Equation 34.

VI-Q)tifar E KQ,PRRL = x 100% (34)

(0.48X500mrem/yr)

Where:

-QL = The maximum site boundary dispersion factor basedon 5 year averages from Table 10.2.

K = Whole body dose factor (mrem/yr per gCi/m3) forisotope "i." Dose factors are presented inTable 10.3.

= Total release rate of isotope "i" from all sources

discharged through this release point in iLCi/sec.

0.48 = Plant vent allocation factor for one unit fromTable 6.2.

500 mrem/yr = Site noble gas dose rate limit.

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The Expected Reading (ER) is the anticipated monitor response based uponthe known plant vent concentration and the monitor response factors. TheExpected Readings for RE-22, RE-44, RE-14 are given by Equations 35, 36,and 37.

ER(E-22)= BKG(+E 22)+k(E22),C(P22), (35)

Where:

ER(RE-22) = Expected reading on monitor RE-22 in pCi/cc.

BKGE -A) = Monitor background.

k(RE22)i = Noble gas monitor response factor for nuclide "i" formonitor RE-22.

C(RE-22) = Concentration of nuclide "i" seen by RE-22.

ER( 44) = BKG(PE44) + CCSP(PE-4 4 )Z k(RE-4 4)IC(PE44), (36)

Where:

ER(RE-4) = Expected reading on monitor RE-44 in ttCi/cc.

BKG(uA4) = Monitor background.

CCSP(REA4) = Conversion constant setpoint for monitor RE-44.

k(RE ), = Noble gas monitor response factor for nuclide "i" formonitor RE-44.

C(RE44), Concentration of nuclide "i" seen by monitor RE-44.

ER(RE 4) =BKG(R 14) +CCSP(RE_14 )Z k(,E 14), C(P 14), (37)

Where:

ER(RE-14) =

BKG(RE-14) =

CCSP(RE-14) =

Expected reading on monitor RE-14 in pCi/cc.

Monitor background.

Conversion constant setpoint for monitor RE-14.

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k-14) = Noble gas monitor response factor for nuclide "i" formonitor RE-14.

C(RE-14)i = Concentration of nuclide "i" seen by monitor RE-14.

Generally if the Expected Reading (ER) is greater than the existing HASPsetting (an "administrative limit" as set by CY2.DCI) then no release shouldbe made until a calculation shows that the HASP (Admin Limit) can beraised so the release can be legally discharged. On the other hand should theER be less than the existing HASP (Admin Limit), then the release can bedischarged.

6.2.7 IPT - PRRL

The Percent Release Rate Limit (PRRL) for radioiodines, tritium andparticulates for each unit is calculated based upon the 1500 mrem/yr organdose rate limit. The dose rate is calculated for the inhalation pathway to thechild age group using the highest (worst case) organ dose factor for nuclide.The Percent Release Rate Limit based on the worst case organ is given byEquation 38.

PRRL0 = I x100% (38)(0.48Xl500mrem/yr)

Where:

V/Qla= = The maximum site boundary dispersion factor based on 5year averages from Table 10.2.

PW = Inhalation dose factor for nuclide "i" (mrem/yr/gCi/m3 ) forchild age group for worst case organ, from Table 10.4.Dose factors are based upon NUREG 0133 methodology.Inhalation dose conversion factors are taken from Reg.Guide 1.109, Rev 1, Table E-9, with the followingexceptions: H-3, Sb-124 and Sb-125 inhalation doseconversion factors taken from NUREG/CR-4013.

Q. = Release rate of isotope "i" in IiCi/sec.

0.48 = Plant vent location factor for one unit from Table 6.2.

1500 mremlyr = Site radioiodine, tritium and particulate dose rate limit.

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6.2.8 Alternate Dose Methodologies

For purposes of routine gaseous effluent dose assessment, the methodologyof NUREG 0133 (described in Section 6.2.5) will be used. However, DCPPmay elect to utilize the dose methodologies of Regulatory Guide 1.109 or theGASPAR computer code for special purposes such as evaluation of potentialnew gaseous effluent dose pathways or critical receptors.

6.2.9 Gas Effluent Dose Projection

The projected dose contributions from each reactor unit due to gaseouseffluents for the current calendar month, quarter and current calendar yearmust be determined in accordance with the methodology and parameters inthe ODCP at least every 31 days.

The computer program, Radioactive Effluent Management System (REMS),is used for this projection. Therefore, by the first day of the year, thefollowing tables in REMS need to be updated:

* GRW dose receptor

* GRW dose rate receptor

* GRW external dose select

* GRW external occupancy

* GRW internal dose select

* GRW internal occupancy

The purpose of this is to determine if appropriate treatment of gaseousradioactive materials in relation to maintaining releases "as low asreasonably achievable," is necessary.

Projections will be made, at least by the end of each month with attention tothe frequency requirement contained in radiological effluent controlsprogram.

The projected dose from each reactor unit is given by:

Dp =Dpu +-Dp~C. (39)

Where:

Dp = Projected Dose.

Dpu = Projected dose attributed to reactor unit, U.

DP.Com Projected dose common to both reactor units.

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The 31 day projected dose is calculated by Equation 40.

DP"' + dc'+ dpBDp =31 x A +A) (40)

Where:

Dp' = Projected Dose for the next 31 day period.

DAP = Previous Month's Actual Dose

d"M = Current Month Actual Dose to date

dCB = Projected Dose from Current Batch Release

T = Number of days in the previous month

t Number of days into the present month

Projected quarterly doses are determined by Equation 41.

D dcQ +(92 +(92 +d-Q +tAB (41)D QA + t (T +t)

Where:

DcQ = Projected dose for the current calendar quarter.

dQ = Current quarter to date actual dose.

DAQ = Previous quarter's actual dose.

dpB = Projected dose as a result of the current batch release.

T = Number of days in the previous quarter.

t = Number of days into the present quarter.

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Projected yearly doses are determined by Equation 42.

Dcy =dcy +(366 - D +d +dB (42)A(T+ ) (2

Where:

Dcy = Projected dose for the current calendar year.

dcy = Current year to date actual dose.

DY = Previous year's actual dose.

d CB = Projected dose as a result of the current batch release.

T = Number of days in the previous year.

t = Number of days into the present year.

6.2.10 Unplanned Gaseous Releases (Abnormal Releases)

a. An unplanned release is an unexpected and potentially unmonitoredrelease to the environment due to operational error or equipmentmalfunctions.

1. Unmonitored unplanned releases shall have a report written bythe Radiochemistry Effluents Engineer describing the event with acalculation, if possible, of the percent of RECP limit. This willthen be forwarded to PSRC for review. Describe these unplannedreleases in the Annual Radioactive Effluent Release Report.

2. Monitored unplanned releases which exceed 1 % of the RECPlimit will also have a report written describing the event and mustbe forwarded to the PSRC for review. Describe these unplannedreleases in the Annual Radioactive Effluent Release Report. Forpurposes of classification only, unplanned release puffs throughthe plant vent may use one hour integrated resolution times.

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6.3 40 CFR 190 Dose Calculations

6.3.1 Pathways

Calculation of total uranium fuel cycle dose for purposes of demonstratingcompliance with 40 CFR 190 requires the contributions from liquid andgaseous effluent as well as direct radiation from the units and outside storagetanks to be considered. The total uranium fuel cycle dose to any member ofthe public will be calculated by summing the following doses:

* Direct Radiation Dose

* Liquid Effluent Dose

* Noble Gas Dose

* Radioiodine, Tritium and Particulate Gaseous Effluent Dose

6.3.2 Methodology

a. Direct Radiation Dose

Determination of direct radiation dose from the reactor units and fromoutside storage tanks may be made by direct survey measurements,derived from environmental TLD data, or calculated by shielding code.

The direct radiation dose will also take into account residence times nearthe site based upon land use census information.

The direct radiation determination using environmental TLD is given byequation 43.

D sb. = [8 ] Dr' x (9.57)x e- (6.38) (43)[8001 0

where:

D's b = the dose rate at the site boundary, in mrem

D'ro = the dose rate from the dosimetry reading, in mrem

r,,d, = the distance from the point source to the dosimetry, in meter

800 = the distance from the point source to site boundary, in meter

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b. Noble Gas Dose

The noble gas skin dose and total body dose contributions to the totaluranium fuel cycle dose to a member of the public will be determined asshown in Equations 44 and 45.

Noble Gas Total Body Dose = 3.17 x 10-8 (x/Q)R KLIQI (44)

Noble Gas Skin Dose =3.17 x 1o8 0 /Q)R (L, +l.1M 1 Y2, (45)

Where:

3.17x10-8 = Conversion constant yr/sec.

Q)R = Maximum historical dispersion factor forreceptor of interest, based on 5 year averagesfrom Table 10.2.

K = Whole body dose factor for nuclide i,-inmrem/yr per pCi/m3. Values are listed inTable 6.3.

Li = Skin dose factor for nuclide i, in mrem/yr perpCi/m3. Values are listed in Table 6.3.

1.1 = Conversion factor mrem/mrad. Converts airdose to skin dose.

M = Gamma air dose factor for nuclide i, in mrad/yrper iLCi/m 3 . Values are listed in Table 10.3.

Q = Total release of noble gas radionuclide, i, in

,uCi/sec.

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c. Liquid and Gaseous Effluent Dose

The doses from liquid effluents and radioiodines, tritium and particulatesin gaseous effluents will be determined by Equations 1 and 33,respectively.

For purposes of calculating the dose required by the radiological effluentcontrols, more realistic assumptions concerning the liquid and gaseouseffluent dose pathways will be used, based upon the most recent land usecensus data as well as the latest environmental monitoring information.

These assumptions may include, but not be limited to: more realisticliquid dilution factors, location and age of actual individuals, sitespecific food pathway parameters, and documentation of true foodconsumption. Other assumptions may be used provided they can besubstantiated by census or direct measurement.

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6.4 On-Site Dose to Members of the Public

Members of the public are occasionally granted access within the site boundary, but onlyin the owner controlled area up to the protected area boundary. The most common publicaccess activities are: tours to the simulator (training building) or Bio Lab, policemenusing the shooting range (most frequent activity), cattle drives through to adjacentproperties, and visits of American Indians to on-site burial grounds (closest to the plant).

Exposure to members of the public due to liquid releases while on-site is highly unlikelyand therefore not addressed. Exposure due to gaseous releases and direct radiation arecredible and therefore are considered.

The dose to members of the public during on-site activities will be primarily determinedby the duration of the on-site visitation time and by the closest proximity to the plant.

For gaseous releases the doses are calculated using Equations 44, 45 and 33. The R1's inEquation 33 consider only the inhalation and ground plane pathway and exclude theinfant age group.

The X/Q and D/Q values are modified using logarithmic extrapolation from the siteboundary to the on-site location of interest as shown in Equations 46 and 47.

I og[X/Q on-sile

1gdistS.B. )- log(dit I/oc0 ogalist-on - site)- log(dist.S.B.)](46)log(ditSB)

+ log[X/QI B

1og[D1Q]onsate ;Z

log[D/Q.B - 1og[D/QJ0 , [log.,,. on - site)- log(dist.SB.)](47)log(dist.S.B.)- log(distloc.) 't

+ log[D/IQ B.

Based upon Regulatory Guide 1.11 1, these equations can be expected to providereasonable dispersion and deposition estimates for distances as close as 200 meters.

Determination of direct radiation dose from the reactor units and from outside storagetanks may be made by direct survey measurements, derived from environmental TLDdata, or calculated by shielding code.

A distance of 200 meters from the plant (both units) equidistant from the plant vent isarbitrarily selected as the closest perimeter for which on-site doses will be calculated.

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The activities of the members-of-the-public while on-site (described above), are at orbeyond 200 meters. Table 6. 3 details the types of on-site activities thatmembers-of-the-public might be expected to participate in at DCPP. The sectors andclosest distances in which they may visit as wvell as expected visitation duration are alsoshown (based on Security Section information).

Table 6. 3

Expected On-Site Distances and Visitation Times for Members of the Public

SECTOR CLOSEST POINT AVERAGE EXPECTEDONSITE MEMBER OF OF APPROACH VISITATIONOF THE PUBLIC VISITATION TO PLANT TIME PER YEAR

Police at SE 700m 208 hoursshooting range

Tour Participants

Simulator Bldg S (SE) 310m 4 hours

Bio Lab SSE (SE) -- 460m 6 hours

Overlook E 210m 1 hours

American Indians NW 200m 96 hoursat burial grounds NNW 200m 96 hours

Ranch hands driving NW 250m 1 hourcattle around site NNW 350m 1 hour

N 320m 1 hourNNE 450m 1 hourNE 630m 1 hour

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7. ACCEPTANCE CRITERIA

7.1 There is no quantitative acceptance for this procedure. If the task or analysis has beenaccomplished within the bounds of this procedure, it is considered acceptable.

8. REFERENCES

8.1 Draft Radiological Tech Specs for PWRs, NUREG No. 0472, May 1978.

8.2 Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents forthe Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, RegulatoryGuide 1.109, Rev. 0, March 1976.

8.3 Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents forthe Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, RegulatoryGuide 1.109, Rev. 1, October 1977.

8.4 Preparation of Radiological Effluent Tech Specs for Nuclear Power Plants,NUREG No. 0133, October 1978.

8.5 LADTAP II - Technical Reference and User Guide, NUREG/CR-4013.

8.6 Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel CycleStandard 40 CFR 190, NUREG No. 0543, January 1980.

'8.7 Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents inRoutine Releases from Light-Water-Cooled Reactors, Regulatory Guide, 1.111,Rev. 1, July, 1977.

8.8 Radioactive Decay Data Tables, David C. Kocher. DOE/TIC-11026, 1981.

8.9 CAP A-6, "Gaseous Radwaste Discharge Management."

8.10 CAP A-5, "Liquid Radwaste Discharge Management."

8.11 CAP D-15, "Steam Generator Leak Rate Determination."

8.12 CAP D-19, "Correlation of Rad Monitors to Radioactivity."

8.13 CY2.DC1, "Radiation Monitoring System High Alarm Setpoint Control Procedure."

8.14 CY2.ID1, "Radiological Effluent and Controls Program" (RECP)

8.15 "Setpoint Calculation for Containment Ventilation Exhaust Monitor,"Calc # NSP-1&2-39-44, 10/92 and 11/92 and AR A0430610.

8.16 NUREG 2919, Computer Code XOQDOQ, Revision 2, September, 1982.

9. RECORDS

9.1 Data Sheets and records will be maintained in the Records Management System (RMS)in accordance with CY1.DC1, "Analytical Data Processing Responsibilities."

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10. APPENDICES

10.1 Tables

11. ATTACHMENTS

11.1 "Liquid Discharges (LRW) Monitored for Radioactivity," 10/04/00

11.2 "Gaseous Releases (GRW) Monitored for Radioactivity," 10/31/00

12. SPONSOR

John Knemeyer

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1 AND 2TITLE: Off-Site Dose Calculations UNITS

TABLE 10.1

LRW COMPOSITE DOSE FACTORS', A..,FOR ADULTS AT A SALTWATER SITE

(mrem/hour per gCi/ml)organ n"o

Nuclide Tot Body Thyroid Kidney Lung GI-LLI Bone Liver

H-3Na-24Cr-51Mn-54Mn-56Fe-55Fe-59Co-57Co-58Co-60Ni-65Cu-64Zn-65Zn-69As-76Br-82Br-84Rb-86Rb-88Rb-89Sr-89Sr-90Sr-91Sr-92Y-90Y-9lmY-92Zr-95Zr-97Nb-95Mo-99Tc-101Ru-103Ru-105Ru-106Ag- 10mSn-113Sn-1 17mSb-122Sb-124Sb-125

1.61E-014.57E-015.58E+001.35E+033.15E+018.23E+037.27E+042.36E+021.35E+033.82E+031.20E+011.OlE+022.32E1+054.56E1+014.42E+014.07E+009.39E-02

2.91E+029.49E-018.34E-01

1.43E+022.83E+033.71E+001.511E+001.63E-012.22E-031.56E-02

3.46E+008.13E-02

1.34E+022.43E+011.88E-01

4.60E+013.51E+002.01E+028.60E+023.53E+038.76E+026.65E+001.09E+024.20E+01

1.611E-014.57E-013.34E+000.OOE+000.OOE+000.OOE+000.OOE+000.00E+000.OOE+000.OOE+000.OOE+000.OOE+000.001E+000.00E + 000.00E+000.OOE+000.OOE+000.OOE+000.00E+000.00E+000.OOE+000.OOE+000.OOE+000.00E+00O.OOE+000.OOE+000.OOE+000.00E+000.00E+000.00E+00

.00E+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+009.85E+022.52E+023.09E-016.70E-011.79E-01

1.61E-014.57E-011.23E+002.101E+032.26E+020.OOE+000.OOE+000.00E+000.OOE+000.OOE+000.00E+005.40E+023.43E+054.26E+028.72E+010.OOE+000.OOE+000.OOE+000.00E+000.OOE+000.OOE+000.00E+000.00E+000.OOE+000.OOE+000.OOE+000.00E+008.02E+002.68E-012.46E+022.89E+023.46E-014.07E+021.15E+023.06E+032.85E+030.00E+000.00E+000.OOE+000.00E+000.OOE+00

1.61E-014.57E-017.40E+000.OOE+000.00E+001.97E+045.30E+040.OOE+000.00E+000.OOE+000.00E+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.00E+000.OOE+000.00E+000.00E+000.OOE+000.00E+000.OOE+000.00E+000.OOE+000.00E+000.00E+000.00E+009.81E-030.00E+000.OOE+000.00E+000.00E+000.00E+000.OOE+000.00E+001.18E+012.15E+021.36E+02

1.61E-014.57E-011.40E+032.16E1+045.67E+032.03E+046.32E+053.59E+031.22E+043.25E+046.65E+021.83E+043.23E+059.85E+010.00E+004.67E+007.37E-071.23E+022.47E-1 16.89E-14

8.OOE+023.55E+034.37E+026.90E+026.42E+041.68E-01

9.32E+031.62E+045.51E+041.51E+062.96E+025.77E-141.25E+045.44E+031.03E+055.911E+050.00E+000.00E+000.00E+007.84E+031.94E+03

0.00E+004.57E-010.00E+000.00E+000.00E+005.1 lE+048.06E+040.00E+000.00E+000.OOE+002.02E+020.OOE+001.61E+053.43E+020.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+004.99E+031.41E+059.18E+013.48E+016.06E+005.73E-025.32E-011.59E+018.81E-014.47E+020.OOE+001.33E-02

1.07E+028.89E+001.59E+031.56E+036.06E+043.02E+032.19E+012.76E+021.77E+02

1.61E-014.57E-01

0.OOE+007.06E+031.78E+023.53E+041.90E+051.42E+026.03E+021.73E+032.62E+012.14E+025.13E+056.56E+024.62E+010.00E+000.00E+006.24E+021.79E+001. 19E + 000.00E+000.OOE+000.00E+000.OOE+000.00E+000.OOE+000.OOE+005.11E+001.78E-01

2.49E+021.28E+021.92E-02

0.00E+000.00E+000.OOE+001.45E+031.66E+033.411E+024.47E-015.22E+001.97E+00

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LRW COMPOSITE DOSE FACTORS', AN,FOR ADULTS AT A SALTWATER SITE

(mrem/hour per ,uCi/ml)organ "o"

Nuclide

Te-129mTe-129Te-13 lmTe-131Te-132I-131I-132I-133I-134I-135Cs-134Cs-136Cs-137Cs-138Ba-139Ba-140Ba-141La-140La-142Ce-141Ce-143Ce-144Pr-144Nd-147W-187Np-239

Tot Body Thyroid Kidney

1.47E+026.19E-01

5.71E+015.03E-011.24E+021.79E+029.96E+OO3.95E+O15.40E+0O2.24E+011.33E+042.04E+037.85E+035.94E+0O2.30E-011.08E+021.29E-012. 1OE-0l9.13E-032.63E-014.94E-02

9.59E+0O9.64E-042.74E-01

2.68E+001.91E-03

3.20E+021.95E+001.08E+021.31E+001.46E+021.02E+059.96E+021.90E+042.62E+024.01E+03O.OOE+0OO.OOE+0OO.OOE+0OO.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+00O.OOE+000.OOE+000.00E+000.00E+00O.OOE+000.OOE+00O.OOE+000.OOE+00

3.89E+031.07E+016.94E+026.99E+001.27E+035.35E+024.54E+012.26E+022.40E+019.75E+015.27E+031.57E+034.07E+038.81E+005.23E-037.02E-012.68E-03

0.OOE+000.OOE+001.OSE+001.97E-01

4.43E+014.44E-032.68E+00O.OOE+001.08E-02

0.0.0.0.0.0.0.0.0.0.1.2.1.831.1

0.0.0.0.0.0.0.-0.0.

Lung GI-LLI Bone LiverOOE+00 4.69E+03 9.31E+02 3.47E+02OOE+00 1.92E+00 2.54E+00 9.55E-01OOE+0O 6.80E+03 1.40E+02 6.85E+01OOE+00 2.26E-01 1.59E+00 6.66E-01OOE+00 6.24E+03 2.04E+02 1.32E+02OOE+00 8.23E+01 2.18E+02 3.12E+02OOE+00 5.35E+00 1.06E+01 2.85E+01OOE+00 1.16E+02 7.45E+01 1.30E+02OOE+00 1.32E-02 5.56E+00 1.51E+01OOE+00 6.87E+01 2.32E+01 6.08E+0175E+03 2.85E+02 6.84E+03 1.63E+0416E+02 3.21E+02 7.16E+02 2.83E+0335E+03 2.32E+02 8.77E+03 1.20E+04.70E-01 5.12E-05 6.07E+00 1.20E+01.17E-03 1.39E+01 7.85E+00 5.59E-0318E+00 3.38E+03 1.64E+03 2.06E+00.63E-03 1.80E-09 3.81E+00 2.88E-03OOE+00 5.83E+04 1.57E+00 7.94E-01OOE+00 2.68E+02 8.06E-02 3.67E-02OOE+00 8.86E+03 3.43E+00 2.32E+00OOE+00 1.67E+04 6.04E-01 4.46E+02OOE+00 6.04E+04 1.79E+02 7.47E+01OOE+00 2.73E-09 1.90E-02 7.87E-03OOE+00 2.20E+04 3.96E+00-4.58E+00OOE+00 2.51E+03 9.16E+00 7.66E+00OOE+00 7.11E+02 3.53E-02 3.47E-03

' Dose factors are based upon NUREG 0133 methodology.

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TABLE 10.2SUMMARY OF LAND USE CENSUS EVALUATION

Sector Receptor Distance X/Q D!Q CommentsDescription (miles)

N none no receptors within 5 miles

NNE residence 4.4 5.70E-8 9.40E-1 1 unknown use

residence 3.3 8.70E-8 1.60E-10 cabin - limited use

NE residence 4.9 3.70E-8 6.20E-11 full time occupancy

ENE residence 4.4 4. 1OE-8 7.60E-1 1 full time occupancy

residence 5.0 3.40E-8 6.OOE-1 1 full time occupancy

E residence 4.0 5.40E-8 1.40E-10 cabin - unknown use

residence 3.7 6.10E-8 1.60E-10 part time occupancy

residence + 4.5 4.40E-8 1.10E-10 full time occupancygarden

ESE oat hay and 3.3 1.50E-7 9.20E-10 field workers present only during thesugar peas day - critical receptor ground plane,

inhalation, and vegetation ingestiondose assessed at this location

SE none no receptors within 5 miles

SSE none over water

S none over water

SSW none over water

SW none over water

WSW none over water

W none over water

WNW none over water

NW highest site 0.5 5.50E-6 1.90E-8 Gas effluent dose rates. PRRLs andboundary HASPs evaluated at this location.

dispersion value

residence 1.2 1.20E-6 4.30E-9 cabin - very limited use

residence 3.6 2.20E-7 6.40E-10 full time occupancy

NNW residence 1.5 7.40E-7 2.20E-9 full time occupancy (trailer) - criticalreceptor ground plane and inhalationdose assessed at this location

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TABLE 10.3

GRW DOSE FACTORS FOR NOBLE GASES'

Whole BodyDose Factor

K,(mrem/yr per pCi/m3 )

Skin Dose FactorLi

(mrem/yr per plCi/m3)

Gamma AirDose Factor

Ma(mrad/yr per plCi/M3)

Beta AirDose Factor

N.(mrad/yr per pCi/m3)Radionuclide

-

Kr-83mKr-85mKr-85Kr-87Kr-88Kr-89Kr-90Xe-131mXe-133mXe-133Xe-135mXe-135Xe-137Xe-138Ar-41

7.56E-021.17E+031.61E+015.92E+031.47E+041.66E+041.56E+049.15E+012.51E+022.94E+023.12E+031.81E+031.42E+038.83E+038.84E+03

1.46E+031.34E+039.73E+032.37E+031.O1E+047.29E+034.76E+029.94E+023.06E+027.1 IE+021.86E+031.22E+044.13E+032.69E+03

1 .93E+011.23E+031.72E+016.17E+031.52E+041.73E+041.63E+041.56E+023.27E+023.53E+023.36E+031.92E+031.51E+039.21E+039.30E+03

2.88E+021.97E+031.95E+031.03E+042.93E+031.06E+047.83E+031.1 lE+031.48E+031.05E+037.39E +022.46E+031.27E+044.75E+033.28E+03

-

' From Table B-I of Regulatory Guide 1.109 (Rev. 1, Oct. 1977)

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1 AND 2TITLE: Off-Site Dose Calculations UNITS

TABLE 10.4CHILD INHALATION PATHWAY DOSE FACTORS FOR WORST CASE ORGAN

RadionuclideH-3CR-51MN-54FE-59CO-58CO-60ZN-65RB-86SR-89Y-90SR-90ZR-95NB-95RU-103RU-106AG-110MSB-124SB-125TE-129MI-1311-133CS-134CS-136CS-137BA-140CE-141CE-144ND-147

PW

6.40E+021.70E+041.58E1+061.27E+061.11 E+067.07E+069.95E+051.98E+052.16E+062.68E+053.85E+072.23E+066.14E+056.62E+051.43E+075.48E+063.24E+062.32E+061.76E+061.62E1+073.85E+061 .O1E+061.711E+059.07E+051.74E+065.44E+051.20E+073.28E+05

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TABLE 10.5

GROUND PLANE DOSE FACTORS

GRW DOSE PARAMETERS', RO GP FOR RADIOIODINES, RADIOACTIVE

PARTICULATES, AND ANY RADIONUCLIDE OTHER THANNOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW),

ANY AGE GROUP, GROUND PLANE PATHWAY(mrem/yr per [LCi/(sec m2 ))

Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLIH-3 0.00E+00 0.00E+00 0.OOE+0+ .00 0.00 O.OOE.0E+00 0.00 .OOE+00CR-51 4.65E+06 4.65E+06 4.65E+06 4.65E+06 4.65E+06 4.65E+06 4.65E+06MN-54 1.38E+09 1.38E+09 1.38E+09 1.38E+09 1.38E+09 1.38E+09 1.38E+09FE-59 2.73E+08 2.73E+08 2.73E+08 2.73E+08 2.73E+08 2.73E+08 2.73E+08CO-58 3.80E+08 3.80E+08 3.80E+08 3.80E+08 3.80E+08 3.80E+08 3.80E+08CO-60 2.15E+10 2.15E+10 2.15E+10 2.15E+10 2.15E+10 2.15E+10 2.15E+10ZN-65 7.46E+08 7.46E+08 7.46E+08 7.46E+08 7.46E+08 7.46E+08 7.46E+08RB-86 8.98E+06 8.98E+06 8.98E+06 8.98E+06 8.98E+06 8.98E+06 8.98E+06SR-89 2.16E+04 2.16E+04 2.16E+04 2.16E+04 2.16E+04 2.16E+04 2.16E+04Y-90 4.50E+03 4.50E+03 4.50E+03 4.50E+03 4.50E+03 4.50E+03 4.50E+03SR-90 O.OOE+00 0.OOE+00 0.00E+00 0.00E+00 O.OOE+00 0.00E+00 0.00E+00ZR-95 2.45E+08 2.45E+08 2.45E+08 2.45E+08 2.45E+08 2.45E+08 2.45E+08NB-95 1.37E+08 1.37E+08 1.37E+08 1.37E+08 1.37E+08 1.37E+08 1.37E+08RU-103 1.08E+08 1.08E+08 1.08E+08 1.08E+08 1.08E+08 1.08E+08 1.08E+08RU-106 4.20E+08 4.20E+08 4.20E+08 4.20E+08 4.20E+08 4.20E+08 4.20E+08AG-llOM 3.45E+09 3.45E+09 3.45E+09 3.45E+09 3.45E+09 3.45E+09 3.45E+09SB-124 5.99E+08 5.99E+08 5.99E+08 5.99E+08 5.99E+08 5.99E+08 5.99E+08SB-125 2.34E+09 2.34E+09 2.34E+09 2.34E+09 2.34E+09 2.34E+09 2.34E+09TE-129M 1.98E+07 1.98E+07 1.98E+07 1.98E+07 1.98E+07 1.98E+07 1.98E+071-131 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+071-133 2.45E+06 2.45E+06 2.45E+06 2.45E+06 2.45E+06 2.45E+06 2.45E+06CS-134 6.90E+09 6.90E+09 6.90E+09 6.90E+09 6.90E+09 6.90E+09 6.90E+09CS-136 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08CS-137 1.03E+10 1.03E+ 10 1.03E+10 1.03E+ 10 1.03E+10 1.03E+10 1.03E+ 10BA-140 2.05E+07 2.05E+07 2.05E+07 2.05E+07 2.05E+07 2.05E+07 2.05E+07CE-141 1.37E+07 1.37E+07 1.37E+07 1.37E+07 1.37E+07 1.37E+07 1.37E+07CE-144 6.96E+07 6.96E+07 6.96E+07 6.96E+07 6.96E+07 6.96E+07 6.96E+07ND-147 8.39E+06 8.39E+06 8.39E+06 8.39E+06 8.39E+06 8.39E+06 8.39E+06


00692326 DOC 08 0604.0656

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NUMBERREVISIONPAGE

CAP A-82655 OF 62


TABLE 10.6

GRW DOSE PARAMETERS FOR RADIOIODINES, RADIOACTIVE PARTICULATES, ANDANY RADIONUCLIDE OTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW)

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PAGE 56 OF 62


TABLE 10.6.1

GRW DOSE PARAMETERS' FOR RADIQIODINES, RADIOACTIVE PARTICULATES, AND ANYRADIONUCLIDE OTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), INFANT

AGE GROUP, INHALATION PATHWAY ORGAN "O" (mrem/yr per gCi/m3 ) R,jnhat

NuclideH-3CR-51MN-54FE-59CO-58CO-60ZN-65RB-86SR-89Y-90SR-90ZR-95NB-95RU-103RU-106AG-110MSB-124SB-125TE-129MI-131I-133CS-134CS-136CS-137BA-140CE-141CE-144ND-147

Bone0.00E+000.00E+000.00E+001.36E+04O.OOE+000.00E+001.93E+040.00E+003.98E+053.29E+031.55E+071.15E+051.57E+042.02E+038.68E+049.98E+033.79E+045.17E+041.41E+043.79E+041.32E+043.96E+054.83E+045.49E+055.60E+042.77E+043. 19E+067.94E+03

Liver3.68E+020.00E+002.53E+042.35E+041.22E+038.02E+036.26E+041.90E+050.00E+000.00E +000.00E+002.79E+046.43E+030.00E+000.00E +007.22E+035.56E+024.77E+026.09E+034.44E+041.92E+047.03E+051.35E+056.12E+055.60E+011.67E+041.21E+068.13E+03

T Body3.68E+028.95E+014.98E+039.48E+031.82E+031.18E+043.11 E+048.82E+041. 14E + 048.82E+013.12E+052.03E+043.78E+036.79E+021.09E+045.OOE+031.20E+041.09E+042.23E+031.96E+045.60E+037.45E+045.29E+044.55E+042.90E+031.99E+031.76E+055.OOE+02

Thyroid3.68E+025.75E+010.OOE+000.OOE+000.OOE+00-0.OOE+000.OOE+000.00E+000.00E+000.00E+O000.OOE+000.00E+000.OOE+000.OOE+000.OOE+000.00E+001.O1E+026.23E+015.47E+031.48E+073.56E+060.00E+000.00E+000.00E+000.OOE+000.OOE+000.OOE+000.00E+00

Kidney3.68E+021.32E+014.98E+030.OOE+000.00E+000.00E+003.25E+040.OOE+000.00E+000.00E+000.00E+003.11 E+044.72E+034.24E+031.07E+051.09E+040.OOE+000.00E+003.18E+045.18E+042.24E+041.90E+055.64E+041.72E+051.34E+015.25E+035.38E+053.15E+03

Lung3.68E+021.28E+041.OOE+061.02E+067.77E+054.51E+066.47E+050.OOE+002.03E+062.69E+051. 12E + 071.75E+064.79E+055.52E+051.16E + 073.67E+062.65E+061.64E+061.68E+060.OOE+000.OOE+007.97E+041.18E+047.13E+041.60E+065.17E+059.84E+063.22E+05

-

GI-LLI3.68E+023.57E+027.06E+032.48E+041.1 1E+043.19E+045.14E+043.04E+036.40E+041.04E+051.31E+052.17E+041.27E+041.61E+041.64E+053.30E+045.91E+041.47E+046.90E+041.06E+032.16E+031.33E+031.43E+031.33E+033.84E+042.16E+041.48E+053.12E+04


00692326.DOC 08 0604.0656

_ _-WiU -*'-

>*-*1UBOSR~tED'POCDUE ^S^O OTUS~TOPEFO~uORKor1SUEORUSE$:-PACIFIC GAS AND ELECTRIC COMPANYDIABLO CANYON POWER PLANT

TITLE: Off-Site Dose Calculations

NUMBER CAP A-8REVISION 26PAGE 57 OF 62

UNITS 1 AND 2

TABLE 10.6.2

GRW DOSE PARAMETERS' FOR RADIOIODINES, RADIOACTIVE PARTICULATES, AND ANYRADIONUCLIDE OTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), CHILD

AGE GROUP, INHALATION PATHWAY ORGAN "O" (mrem/yr per iiCi/m3 ) R, jnhal

Nuclide Bone Liver T Body -Thyroid Kidney Lung GI-LLIH-3 0.00E+00 6.40E+02CR-51 0.00E+00 0.00E+00MN-54 0.00E+00 4.29E+04FE-59 2.07E+04 3.34E+04CO-58 0.00E+00 1.77E+03CO-60 0.OOE+00 1.31E+04ZN-65 4.26E+04 1.13E+05RB-86 0.00E+00 1.98E+05SR-89 5.99E+05 0.00E+00Y-90 4.11E+03 0.00E+00SR-90 3.85E+07 0.00E+00ZR-95 1.90E+05 4.18E+04NB-95 2.35E+04 9.18E+03RU-103 2.79E+03 0.00E+00RU-106 1.36E+05 0.OOE+00AG-llOM 1.69E+04 1.14E+04SB-124 5.74E+04 7.40E+02SB-125 9.84E+04 7.59E+02TE-129M 1.92E+04 6.85E+03I-131 4.81E+04 4.81E+04I-133 1.66E+04 2.03E+04CS-134 6.51E+05 1.01E+06CS-136 6.51E+04 1.71E+05CS-137 9.07E+05 8.25E+05BA-140 7.40E+04 6.48E+01CE-141 3.92E+04 1.95E+04CE-144 6.77E+06 2.12E+06ND-147 1.08E+04 8.73E+03

6.40E+021.54E+029.51E+031.67E1+043.16E+032.26E+047.03E+041. 14E + 051.72E+041.1 lE+027.66E+053.70E+046.55E+031.07E+031.69E+049.14E+032.OOE+042.07E+043.04E+032.73E+047.70E+032.25E+051. 16E + 051.28E+054.33E+032.90E+033.6E1+056.81E+02

6.40E+028.55E1+010.00E+000.00E+000.00E+000.OOE+000.OOE+000.OOE+000.00E+000.00E+000.OOE+000.00E+000.00E+000.00E+000.OOE+000.OOE+001.26E+029.10iE+016.33E+031.62E+073.85E+060.OOE+000.OOE+000.OOE+000.OOE+000.00E +000.OOE+000.OOE+00

6.40E+022.43E+011.OOE+040.OOE+000.00E+000.00E+007.14E+040.OOE+000.00E+000.00E+000.OOE+005.96E+048.62E+037.03E+031.84E+052.12E+040.OOE+000.OOE+005.03E+047.88E+043.38E+043.30E+059.55E+042.82E+052.1 1E+018.55E+031.17E+064.81E+03

6.40E+021.70E+041.58E+061.27E+061.11E+067.07E+069.95E+050.OOE+002.16E+062.62E+051.48E+072.23E+066.14E+056.62E+051.43E+075.48E+063.24E+062.32E+061.76E+060.003E+000.00E+001.21E+051.45E+041.04E+051.74E+065.44E+051.20E+073.28E+05

6.40E+021.08E+032.29E+047.07E+043.44E+049.62E+041.63E+047.99E+031.67E+052.68E+053.43E+056.1 1E+043.70E+044.48E+044.29E+051.OOE+051.64E+054.03E+041.82E+052.84E+035.48E+033.85E+034.18E+033.62E+031.02E+055.66E1+043.89E+058.21E+04

Dose factors are based upon NUREG 0133 methodology.

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PAGE 58 OF 62


TABLE 10.6.3

GRW DOSE PARAMETERS' FOR RADIOIODINES, RADIOACTIVE PARTICULATES, AND ANYRADIONUCLIDE OTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), TEEN AGE

I GROUP, INHALATION PATHWAY ORGAN "O" (mrem/yr per tCilm3 ) R.nhal

NuclideH-3CR-51MN-54FE-59CO-58CO-60ZN-65RB-86SR-89Y-90SR-90ZR-95NB-95RU-103RU-106AG-110MSB-124SB-125TE-129MI-131I-133CS-134CS-136CS-137BA-140CE-141CE-144ND-147

-

- Bone0.00E+000.00E+000.00E+001.59E+040.00E+000.00E+003.86E+040.00E+004.34E+052.98E+033.311E+071.46E+051.86E+04

'2.10E+039.84E+041.38E+044.30E+047.38E+041.39E+043.54E+041.22E+045.02E+055.15E+046.70E+055.47E+042.84E+044.89E+067.86E+03

Liver7.25E+020.OOE+005.11 E+043.70E+042.07E+031.51E+041.34E+051.90E+050.OOE+000.OOE+000.OOE+004.58E+041.03E+040.OOE+000.00E+001.31E+047.94E+028.08E+026.58E+034.91E1+042.05E+041.13E+061.94E+058.48E+056.70E1+011.90E+042.02E1+068.56E+03

T Body7.25E+021.35E+028.40E+031.43E+042.78E+031.98E+046.24E1+048.40E+041.25E+048.OOE+016.66E+053.15E+045.66E+038.96E+021.24E+047.99E+031.68E+041.72E+042.25E+032.64E+046.22E+035.49E+051.37E+053.1 lE+053.52E+032.17E+032.62E+055.13E+02

Thyroid7.25E+027.50E+010.OOE+000.OOE+000.OOE+000.OOE+000.00E+000.OOE+000.OOE+000.00E+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.00E+009.76E1+017.04E1+014.58E1+03

-1.46E+072.92E+060.00E+000.00E+000.00E+000.00E+000.00E+000.00E+000.OOE+00

Kidney7.25E+023.07E+011.27E+040.OOE+000.OOE+000.OOE+008.64E+040.00E+000.OOE+000.OOE+000.OOE+006.74E+041.OOE+047.43E+031.90E+052.50E+040.OOE+000.00E+005.19E+048.40E+043.59E+043.75E+051.101E+053.04E+052.28E+018.88E+031.21E+065.02E+03

Lung - GI-LLI7.25E+022.101E+041.98E+061.53E+061.34E+068.72E+061.24E+060.OOE+002.42E+062.93E+051.65E+072.69E+067.51E+057.83E1+051.61E+076.75E+063.85E+062.74E+061.98E+060.OOE+000.00E+001.46E+051.78E+041.21E+052.03E1+066.14E1+051.34E+073.72E+05

7.25E+023.OOE+036.68E1+041.78E+059.52E+042.59E+054.66E+041.77E+043.71E+055.59E1+057.65E+051.49E+059.68E+041.09E+059.60E+052.73E+053.98E+059.92E+044.05E+056.49E+031.03E+049.76E+031.09E+048.48E+032.29E+051.26E+058.64E+051.82E+05

1 Dose factors are based upon NUREG 0133 methodology.

00692326 DOC 08 0604.0656

t .**UVCONTROLLED PROCEDURE - DO NOT USE TO PERFORM'WORK or ISSUEFOR USE.*KsPACIFIC GAS AND ELECTRIC COMPANYDIABLO CANYON POWER PLANT

NUMBER CAP A-8REVISION 26PAGE 59 OF 62


TABLE 10.6.4

GRW DOSE PARAMETERS' FOR RADIOIODINES, RADIOACTIVE PARTICULATES, AND ANYRADIONUCLIDE OTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), ADULT

AGE GROUP, INHALATION PATHWAY ORGAN "O" (mrem/yr per ,uCi/m3) RInhal

Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLIH-3 0.OOE+00 7.18E+02 7.18E+02 7.18E+02 7.18E+02 7.18E+02 7.18E+02CR-51 0.OOE+00 0.OOE+00 1.00E+02 5.95E+01 2.28E+01 1.44E+04 3.32E+03MN-54 0.00E+00 3.96E+04 6.30E+03 0.OOE+00 9.84E+03 1.40E+06 7.74E+04FE-59 1.18E+04 2.78E+04 1.06E+04 0.OOE+00 0.OOE+00 1.02E+06 1.88E+05CO-58 0.OOE+00 1.58E+03 2.07E+03 0.00E+00 0.OOE+00 9.28E+05 1.06E+05CO-60 0.00E+00 1.15E+04 1.48E+04 0.OOE+00 0.OOE+00 + 5.97E+06 2.85E+05ZN-65 3.24E+04 1.03E+05 4.66E+04 O.OOE+00 6.90E+04 8.64E+05 5.34E+04RB-86 0.OOE+00 1.35E+05 5.90E+04 0.OOE+00 0.00E+00 0.00E+00 1.66E+04SR-89 3.04E+05 0.OOE+00 8.72E+03 0.OOE+00 O.OOE+00 1.40E+06 3.50E+05Y-90 2.09E+03 0.OOE+00 5.61E+01 0.00E+00 0.OOE+00 1.70E+05 5.06E+05SR-90 2.87E+07 O.OOE+00 5.77E+05 0.OOE+00 0.OOE+00 9.60E+06 7.22E+05ZR-95 1.07E+05 3.44E+04 2.33E+04 0.OOE+00 5.42E+04 1.77E+06 1.50E+05NB-95 1.41E+04 7.82E+03 4.21E+03 0.00E+00 7.74E+03 5.05E+05 1.04E+05RU-103 1.53E+03 0.OOE+00 6.58E+02 0.00E+00 5.83E+03 5.05E+05 1.10E+05RU-106 6.91E+04 0.OOE+00 8.72E+03 0.OOE+00 1.34E+05 9.36E+06 9.12E+05AG-llOM 1.08E+04 1.00E+04 5.94E+03 0.OOE+00 1.97E+04 4.63E+06 - 3.02E+05SB-124 3.12E+04 5.89E+02 1.24E+04 7.55E+01 0.OOE+00 2.48E+06 4.06E+05SB-125 5.34E+04 5.95E+02 1.26E+04 5.40E+01 0.OOE+00 1.74E+06 1.01E+05TE-129M 9.76E+03 4.67E+03 1.58E+03 3.44E+03 3.66E+04 1.16E+06 3.83E+051-131 2.52E+04 3.58E+04 2.05E+04 1.19E+07 6.13E+04 0.OOE+00 6.28E+031-133 8.64E+03 1.48E+04 4.52E+03 2.15E+06 2.58E+04 0.OOE+00 8.88E+03CS-134 3.73E+05 8.48E+05 7.28E+05 0.OOE+00 2.87E+05 9.76E+04 1.04E+04CS-136 3.90E+04 1.46E+05 1.10E+05 0.OOE+00 8.56E+04 1.20E+04 1.17E+04CS-137 4.78E+05 6.21E+05 4.28E+05 0.00E+00 2.22E+05 7.52E+04 8.40E+03BA-140 3.90E+04 4.90E+01 2.57E+03 0.OOE+00 1.67E+01 1.27E+06 2.18E+05CE-141 1.99E+04 1.35E+04 1.53E+03 0.00E+00 6.26E+03 3.62E+05 1.20E+05CE-144 3.43E+06 1.43E+06 1.84E+05 0.OOE+00 8.48E+05 -7.78E+06 8.16E+05ND-147 5.27E+03 6.10E+03 3.65E+02 0.00E+00 3.56E+03 2.21E+05 1.73E+05


00692326.DOC 08 0604.0656

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PAGE 60 OF 62


TABLE 10.6.5

GRW DOSE PARAMETERS1 FOR RADIOIODINES, RADIOACTIVE PARTICULATES, AND ANYRADIONUCLIDE OTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), CHILD

AGE GROUP, VEGETATION PATHWAY ORGAN "0" (mrem/yr per tlCi/(sec m2)) R.,Vegr

Nuclide Bone Liver T Body Thyroid Kidney LungH-32

CR-51MN-54FE-59CO-58CO-60ZN-65RB-86SR-89Y-90SR-90ZR-95NB-95RU-103RU-106AG-110MSB-124SB-125TE-129MI-131I-133CS-134CS-136CS-137BA-140CE-141CE-144ND-147

0.OOE+000.OOE+000.OOE+003.97E+080.OOE+000.OOE+008.12E+080.OOE+003.59E+102.31E+041.87E+123.86E+064.12E+051.53E+077.45E+083.21E+073.52E+084.99E+088.40E+081.43E+083.52E1+061.60E+ 108.18E+072.39E+102.77E+086.55E+051.27E+087.27E+04

2.29E+030.OOE+006.65E+086.42E+086.45E+073.78E+082.16E+094.54E+080.O0E+000.OOE+000.OOE+008.50E+051.61E+050.OOE+000.OOE+002.17E+074.57E+063.85E+062.35E+081.44E+084.36E+062.63E+ 102.25E+082.29E+ 102.43E+053.27E+053.98E+075.89E+04

2.29E+031.17E+051.77E+083.20E+081.97E+081.12E+091.35E+092.79E+081.03E+096.18E+023.77E+ 107.56E+051.15E+055.89E+069.30E+071.74E+071.23E+081.05E+081.30E+088.17E+071.65E+065.55E+091.46E+083.38E+091.62E+074.85E+046.78E+064.56E+03

2.29E+036.49E+040.O0E+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+007.78E + 054.62E+052.71E+084.75E+ 108.09E+080.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+00

2.29E+031.77E+041.86E+080.OOE+000.OOE+000.OOE+001.36E+090.OOE+000.OOE+000.OOE+000.OOE+001.22E+061.51E+053.86E+071 .O1E+094.04E+070.OOE+000.OOE+002.47E+092.36E+087.26E+068.16E+091.20E+087.46E+097.90E+041.43E+052.21E+073.23E+04

2.29E+031.18E+050.OOE+001.86E+080.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+000.OOE+001.96E+082.78E+080.00E+000.OOE+000.OOE+002.93E+091.79E+072.68E+091.45E+050.OOE+000.OOE+000.OOE+00

GI-LLI2.29E+036.20E+065.58E+086.69E + 083.76E+082.1OE+093.80E+082.92E+07

- 1.39E+096.57E+07

* 1.67E+ 108.86E+082.97E+083.96E+081. 16E+ 102.58E+092.20E+091.19E+091.02E+091.28E+071.76E+061.42E+087.90E+061.43E+081.40E+084.08E+081.04E+ 109.33E+07


2 For Tritium the units of the dose parameters are mrem/yr per ACi/m 3 for all pathways, and they mustbe multiplied by X/Q.

00692326.DOC 08 0604 0656


PAGE 61 OF 62


TABLE 10.6.6

GRW DOSE PARAMETERS1 FOR RADIOIODINES, RADIOACTIVE PARTICULATES, AND ANYRADIONUCLIDE OTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), TEEN AGE

GROUP, VEGETATION PATHWAY ORGAN "0" (mrem/yr per gCiI(sec in2)) R,,Vg

Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLIH-3 2 0.OOE+00 1.47E+03 1.47E+03 1.47E+03 1.47E+03 1.47E+03 1.47E+03CR-51 0.OOE+00O0.OOE+00 6.16E+04 3.42E+04 1.35E+04 8.79E+04 1.03E3+07MN-54 0.OOE3+00 4.54E+08 9.01E3+07 0.OOE+00 1.36E+08 0.OOE+00 9.32E+08FE-59 1.79E+08 4.18E+08 1.61E+08 0.OOE+00 O.OOE3+00 1.321E+08 9.89E+08CO-58 0.OOE+00 4.37E+07 1.01E+08 0.OOE3+00O0.OOE+00 0.OOE3+00 6.02E+08CO-60 0.OOE+00 2.49E+08 5.60E+08 0.OOE+00O0.OOE+00 0.OOE3+00 3.24E+09ZN-65 4.24E+08 1.47E+09 6.86E+08 0.OOE3+00 9.41E+08 O.OOE+00 6.23E+08RB-86 O.OOE+00 2.75E+08 1.29E+08 0.OOE3+00O0.OOE+00 0.OOE3+00 4.06E+07SR-89 1.51E+10 O.OOE+00 4.33E+08 0.OOE+00 O.OOE+00 O.OOE+00 1.80E+09Y-90 1.24E+04 0.OOE+00 3.35E+02 0.OOE+00O0.OOE+00 0.OOE+00 1.02E+08SR-90 9.22E+l1 0 .OOE+00 1.84E+10 0.OOE1+00 O.OOE+00 0.OOE+00 2.1lE+10ZR-95 1.72E+06 5.44E+05 3.74E+05 0.OOE+00 7.99E+05 0.OOE+00 1.26E+09NB-95 1.93E+05 1.07E+05 5.90E+04 0.OOE+00 1.04E+05 0.OOE3+00 4.58E+08RU-103 6.82E+06 0.OOE+00 2.91E+06 0.OOE+00 2.40E+07 O.OOE+00- 5.69E+08RU-106 3.09E+08 0.OOE+00 3.90E+07 0.OOE+00 5.97E+08 O.OOE3+00 1.48E+l10AG-110M 1.52E+07 l.44E+07 8.73E+06 0.OOE+00 2.74E+07 0.OOE+00 4.03E+09SB-124 1.55E+08 2.85E+06 6.03E+07 3.51E+05 O.OOE+00 1.35E+08 3.11E+09SB-125 2.14E+08 2.34E+06 5.01E+07 2.05E+05 0.OOE+00 1.88E+08 1.67E+09TE-129M 3.61E+08 1.34E+08 5.72E+07 1.17E+08 1.51E+09 0.OOE+00 1.36E+091-131 7.68E+07 1.08E+08 5.78E+07 3.14E+ 10 1.85E+08 0.OOE+00 2.13E+071-133 1.93E+06 3.28E+06 1.OOE+06 4.58E+08 5.75E+06 0.OOE+00 2.48E+06CS-134 7.101E+09 1.67E+lO0 7.75E+09 0.OOE+00 5.31E+09 2.03E+09 2.08E+08CS-136 4.35E+07 1.71E+08 1.15E+08 0.OOE+00 9.31E+07 1.47E+07 1.38E+07CS-137 1.O1E3+10 1.35E+l10 4.69E+09 O.OOE+00 4.59E+09 l.78E+09 1.92E+08BA-140 1.38E+08 1.69E+05 8.90E+06 0.OOE1+00 5.74E+04 l. 14E +05 2.13E+08CE-141 2.83E+05 1.89E+05 2.17E+04 O.OOE+00 8.89E+04 0.OOE3+00 5.40E+08CE-144 5.27E+07 2.18E+07 2.83E+06 0.OOE1+00 1.30E+07 0.OOE3+00 1.33E+10ND-147 3.67E+04 4.OOE+04 2.39E+03 0.OOE+00 2.35E+04 0.OOE+00 1.44E+08

1 Dose factors are based upon NUREG 0133 methodology.

2 For Tritium the units of the dose parameters are mrem/yr per pCi/M3 for all pathways, and they mustbe multiplied by XIQ.

00692326 DOC 08 0604.0656

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PAGE 62 OF 62


TABLE 10.6.7

GRW DOSE PARAMETERS' FOR RADIOIODINES, RADIOACTIVE PARTICULATES, AND ANYRADIONUCLIDE OTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), ADULT

AGE GROUP, VEGETATION PATHWAY ORGAN "O"(mrem/yr per piCi/(sec m2))R veg,

Nuclide Bone Liver T Body Thyroid Kidney I Lung I GI-LLIH-32 0.OOE+00 1.29E+03 1.29E+03 1.29E+03 1.29E+03 1.29E+03 1.29E+03CR-51 0.OOE+00 0.OOE+00 4.64E+04 2.77E+04 1.02E+04 6.15E+04 1.17E+07MN-54 0.OOE+00 3.13E+08 5.97E+07 0.00E+00 9.31E+07 0.00E+00 9.58E+08FE-59 1.26E+08 2.96E+08 1.13E+08 0.00E+00 0.OOE+00 8.27E+07 9.87E+08CO-58 0.00E+00 3.08E+07 6.90E+07 0.00E+00 0.OOE+00 0.00E+00 6.24E+08CO-60 0.00E+00 1.67E+08 3.69E+08 0.00E+00 0.OOE+00 0.00E+00 3.14E+09ZN-65 3.17E+08 1.01E+09 4.56E+08 0.00E+00 6.75E+08 0.OOE+00 6.36E+08RB-86 0.OOE+00 2.20E+08 1.03E+08 0.OOE+00 0.OOE+00 0.00E+00 '4.34E+07SR-89 9.95E+09 0.OOE+00 2.86E+08 0.OOE+00 0.00E+00 0.00E+00 1.60E+09Y-90 1.33E+04 0.OOE+00 3.57E+02 0.00E+00 0.OOE+00 0.00E+00 1.41E+08SR-90ZR-95NB-95RU-103RU-106AG-110MSB-124SB-125TE-129M1-1311-133CS-134CS-136CS-137BA-140CE-141CE-144ND-147

6.95E+ 111.18E+061.43E+054.77E+061.93E+081.05E+071.04E+081.37E+082.51E+088.07E+072.08E+064.67E+094.25E+076.36E+091.29E+081.97E+053.29E+073.37E+04

0.OOE+003.77E+057.95E+040.OOE+000.00E+009.75E+061.96E+061.53E+069.37E+071.15E+083.62E+06-1.11E+101.68E+088.70E+091.61E+051.33E+051.38E+073.90E+04

1.40E+102.55E+054.27E+042.05E+062.44E+075.79E+064.1 1E+073.25E+073.97E+076.62E+071.10E+069.08E+091.21E+085.70E+098.42E+061.51E+041.77E+062.33E+03

0.00E+000.OOE+000.OOE+000.OOE+000.00E+000.OOE+002.52E+051.39E+058.62E+073.78E+105.32E+080.00E+000.OOE+000.00E+000.OOE+000.00E+00O.00E+000.OOE+00

0.OOE+005.92E+057.86E+041.82E+073.72E+081.92E+070.OOE+000.OOE+001.05E+091.98E+086.31E+063.59E+099.33E+072.95E+095.49E+046.19E+048.16E+062.28E+04

0.00E+000.00E+000.00E+000.00E+000.00E+000.00E+008.08E+071.05E+080.OOE+000.00E+000.00E+001.19E + 091.28E+079.81E+089.24E+040.OOE+000.OOE+000.OOE+00

1.75E+101.20E+094.83E+085.57E+081.25E+103.98E+092.95E+091.50E+091.26E+093.05E+073.25E+061.94E+081.90E+071.68E+082.65E+085.09E+08l.11E+101.87E+08


2 For Tritium the units of the dose parameters are mrem/yr per piCi/m 3 for all pathways, and they mustbe multiplied by X/Q.

00692326.DOC 08 0604.0656

* * UNCONTROLLED NT USE TO E WR _ _SSUE FOR.USE. ._ ,

10/04/00 Page 1 of 1DIABLO CANYON POWER PLANT

CAP A-8ATTACHMENT 11.1 1AND s

TITLE: Liquid Discharges (LRW) Monitored for Radioactivity

LIQUID RADWASTE SYSTEM CONDENSATE DEMINERALIZER STEAM GENERATORBLOWDOWN

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00692326.DOC 08 0604.0656

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00692326.DOC 08 0604.0656 00692319


Attachment 4

Process Control Program

(Procedure RP2.DC2, Revisions 11; includes revision 8- 11)

[t.St'*-*.U/COA TROLLEffPROCEDURE *DONOTUSEi'TOPERFORM WORK or ISSUEFOR MR

PACIFIC GAS AND ELECTRIC COMPANY NUMBER RP2.DC2NUCLEAR POWER GENERATION REVISION 11DIABLO CANYON POWER PLANT PAGE 1 OF 6ADMINISTRATIVE PROCEDURE

TITLE: Process Control Program

10/23102EFFECTIVE DATE

PROCEDURE CLASSIFICATION: QUALITY RELATEDSPONSORING ORGANIZATION: RADIATION PROTECTION

REVIEW LEVEL: "A"

1. SCOPE

1.1 The purpose of the Process Control Program (PCP) is to define the necessary programguidance used at the plant to ensure that SOLID RADIOACTIVE WASTEMANAGEMENT activities, in solidifying wet radioactive waste for disposal, conform tothe Code of Federal and State Regulations and the Waste Burial Site License Criteria.

2. DISCUSSION

2.1 Solidification is the conversion of wet radioactive wastes into a form that meets shippingand burial ground requirements.

2.2 This procedure implements the requirements of 10 CFR 50.36a and General DesignCriterion 60 of Appendix A to 10 CFR Part 50. The process parameters included inestablishing the PROCESS CONTROL PROGRAM may include, but are not limited to,waste type, waste pH, waste/liquid/SOLIDIFICATION agent/catalyst ratios, waste oilcontent, waste principal chemical constituents, and mixing and curing times.

2.3 This procedure contains the individual procedures necessary to perform PCP samplesolidifications.

2.4 Cement solidification will not be utilized to stabilize resin orfloor drain sludges. OnlyNRC approved binders, state approved binders or binders submitted for state approval(e.g., Advanced Polymer) may be used to solidify resin or floor drain sludges to meetwaste form stability.

3. RESPONSIBILITIES

3.1 Station director has the overall responsibility for the solid radioactive waste activities andreviews and approves changes to the PCP.

3.2 Radiation protection manager is responsible for the implementation of the requirements ofthis procedure.

3.3 Radwaste engineer is responsible for the development and review of procedures relatingto the requirements of this procedure.

3.4 The radwaste foreman is responsible for the implementation of procedures relating to therequirements of this procedure.

3.5 Quality is responsible for verification of compliance with the program requirements.

011804,/D/OC OIB 1023 0812

//'

fie**=*-UN/CONTROLLED PROCEDURE-tDO NKOT USE To PERFORM WORK or ISSUE FOR USE * * *=rPACIFIC GAS AND ELECTRIC COMPANY NUMBER RP2.DC2DIABLO CANYON POWER PLANT REVISION 11

PAGE 2 OF 6


4. PREREQUISITES

4.1 Changes to this procedure requires submission to the US NRC in the annual RadioactiveEffluent Release report for the period in which the changes were made.

4.2 Any major change to the solid radwaste treatment system shall be reported to the USNRC in the annual Radioactive Effluent Release report for the period in which the changewas approved. The discussion of each change shall contain the items listed inAttachment 7.2. This information may be submitted as part of the annual FSAR update inlieu of the annual Radioactive Effluent Release report.

5. INSTRUCTIONS

5.1 GENERAL

It is the policy of the company to conscientiously apply emphasis and attention to thoseactivities associated with generation, processing, packaging, storage and disposal ofradioactive waste generated at the plant and to maintain a high level of assurance thatradioactive waste products meet or exceed the applicable federal and state regulations andthe radioactive waste burial site license criteria.

5.2 WET WASTE

5.2.1 LIQUID/WET WASTE

Liquid/wet wastes are processed to a condition meeting shipping and disposalcriteria. These criteria include requirements for immobilization, stability andlimits on free standing water (FSW). Specific instructions on processing andrequired FSW limits are contained in plant procedures and/or qualified vendorprocedures.

5.2.2 CONTAINERS, SHIPPING CASKS AND PACKAGING

Solid radioactive waste is processed, packaged and shipped in accordance withplant procedures and/or qualified vendor procedures. These proceduresprovide specific instructions which ensure the container, shipping casks, andpackaging methods comply with the applicable code of federal regulations,state regulations and the radioactive waste burial site license criteria.

5.2.3 SHIPPING AND DISPOSAL

Solid radioactive waste is prepared, loaded and shipped to a federal and/or statelicensed radioactive waste disposal facility (burial ground) in accordance withplant procedures and/or qualified vendor procedures. These proceduresprovide specific instructions which ensure the shipments meet the intendedburial site license requirements as well as applicable federal and stateregulations.

01180412 DOC OIB 1023 0812

FWRaSUNCONTRLEPACIFIC GAS AND ELECTRIC COMPANY NUMBER RP2.DC2DIABLO CANYON POWER PLANT REVISION 11

PAGE 3 OF 6


5.2.4 LABORATORY MIXING OF SAMPLES

Qualified vendor procedures, reviewed and approved by the station director,provide written instructions on sampling, processing and handling waste for thedetermination of process parameters prior to the actual full scale solidification.These procedures contain the description of the laboratory mixing methodsused for these samples.

5.2.5 SOLIDIFICATION PROCESS

Qualified vendors used for radioactive waste solidification are required toprovide the Process Control program and written procedures. Theseprocedures and changes thereto must be reviewed and approved by the stationdirector prior to use. Further, the vendors are required to have an NRC topicalreport, state approval or submittal for state approval on the waste forms whichwill be solidified. These documents should include:

a. Description of the solidification process.

b. Type of solidification used.

c. Process control parameters.

d. Parameter boundary conditions.

e. Proper waste form properties.

f. Specific instructions to ensure the systems are operated withinestablished process parameters.

5.2.6 SAMPLING PROGRAM FOR SOLIDIFICATION

Vendors, utilized for radioactive waste solidification, are required to include intheir approved procedures, requirements to sample at least every tenth batch ofthe same waste type to ensure solidification and to provide actions to be takenif a sample fails to verify solidification. After a test specimen failure, initialtest specimens from three consecutive batches of that waste type mustdemonstrate solidification before testing requirements of every tenth batch canbe resumed. Verification of such sampling is to be accomplished bycompleting Form 69-10350, "Processing Control Program (PCP) Verification."(See Attachment 7.1.) These forms will be maintained by radiation protectionand in the Records Management System (RMS). These procedures andchanges thereto must be reviewed and approved by the station director prior touse.

01180412DOC 01B 10230812

-m eEROR-M !-&SSE OPACIFIC GAS AND ELECTRIC COMPANY NUMBER RP2.DC2DIABLO CANYON POWER PLANT REVISION 11

PAGE 4 OF 6


5.2.7 WASTE FORM VERIFICATION

Vendors utilized to process wet wastes are required to include in theirprocedures provisions to verify that the solidification and/or FSW criteria inthe federal and state regulations and the burial site license are met for thespecific type of waste being processed.

5.2.8 CORRECTIVE ACTIONS FOR FREE STANDING WATER

Vendors utilized to process wet wastes are required to include in theirapproved procedures provisions for correcting processed waste in which freestanding water in excess of the FSW criteria is detected.

5.2.9 EXOTHERMIC PROCESSES

Vendors utilized for radioactive waste solidification that use an exothermicsolidification method are required to include in their approved procedures:

a. Waste/binder temperature monitoring to mitigate the consequence ofadverse exothermic reactions which may occur in the full scalesolidification but might not be noticeable in the specimen tests.

b. Specific process control parameters that shall be met before capping thecontainer.

5.3 OILY WASTE

Oily wastes are shipped to off-site processor for treatment. These processors provide theproper methods to treat oily wastes to comply with federal and state regulations andapplicable burial site license criteria.

5.4 SPECIAL CASES

Based upon previous industry experience, the plant foresees the potential for situationsarising that may be beyond existing plant capabilities. Anticipating this possibility,provisions are made herein to accommodate such situations in a timely manner by usingspecial techniques or processes. These special cases would be controlled as follows:

5.4.1 Implementing procedures would be developed comparable to those used fornormal plant solid waste activities based on the guidance of this PCP andincorporating the applicable provisions for process control and testing.

5.4.2 The implementing procedure would receive station director approval prior touse.

5.4.3 Use of this provision and supporting information would be included in the nextannual Radioactive Effluent Release report to the NRC.

01180412DOC 01B 1023 0812

U't' mUNlCO/VTRO!LLEDPROL7EIURE- 170. NO/UUSE TO PEAFORM WORK or 1/SSUE FOR USE>^= ~*->PACIFIC GAS AND ELECTRIC COMPANY NUMBER RP2.DC2DIABLO CANYON POWER PLANT REVISION 11

PAGE 5 OF 6


5.5 REMEDIAL ACTIONS

5.5.1 For waste forms which do not meet federal, state and burial site regulations andrequirements, suspension of shipment of the inadequately processed waste andcorrection of the PCP, procedures or processing equipment shall be performedas necessary to prevent recurrence.

5.5.2 For waste forms not prepared in accordance with the PCP, testing of the wasteto verify shipping and burial site requirements shall be performed andappropriate administrative action taken to prevent recurrence.

5.6 VENDOR PROCEDURES AND REPORTS

5.6.1 The following are maintained in Document Control Master File, Catalog No.TK 9400/ATG-1.

a. Allied Technology Group, Inc., Procedure PG0I-PCP-05-001, ProcessControl Program for ATG Radwaste Solidification at Diablo Canyon,Rev. I

b. Topical Report TR-002, 10 CFR 61 Qualified Radioactive Waste Forms,Rev. I

c. Allied Technology Group, Inc., Procedure STD-P-01-035, "TemperatureIndicating Device Comparison Test," Rev. 0

d. Allied Technology Group, Inc., Procedure STD-P-01-020, "Calibrationfor a Triple Beam Balance," Rev. 0

e. Allied Technology Group, Inc., Procedure STD-P-01-024, "PenetrometerCheck Procedure," Rev. 0

5.6.2 -The following are maintained in Document Control Master File, Catalog No.TK 94001 DTI-I.

a. Diversified Technologies, Procedure DTS-AP-PCP-20-OPS-01, PCP forAdvanced Polymer Solidification, Rev. E.

b. Diversified Technologies, Procedure DT-DCPP-10, PCP for VERISolidification, Rev 14

c. Topical Report DTI-VERI-1 00-NP-A. VERITr (Vinyl Ester Resin InSitu) Solidification Process for Low-Level Radioactive Waste, Rev I

d. Topical Report DT-VERI-I00-NP-A, Addendum .ENCAPEncapsulation Utilizing the VERI Solidification Process.

e. Diversified Technologies, Procedure QIP 20-17, pH Meter Calibration,Rev. 3

f. Diversified Technologies, Procedure QIP 20-18, Conductivity MeterCalibration, Rev. 2

01180412DOC 01B 10230812

OROL CEDURE-TO P M WORKorISSUEFORSSPACIFIC GAS AND ELECTRIC COMPANY NUMBER RP2.DC2DIABLO CANYON POWER PLANT REVISION 11

PAGE 6 OF 6


g. Diversified Technologies, Procedure QIP-20-16, ThermocoupleCalibration, Rev I

h. Diversified Technologies, Procedure DT-DCPP-VES-14, PCP for VESSolidification, Rev B.

i. Topical Report DNS-RSS-200-NP, The Dow Waste SolidificationProcess for Low-Level Radioactive Waste (Docket Number WM-82).

6. RECORDS

6.1 Records of PCP specimen results and Form 69-10350 shall be submitted to the RecordsManagement System on a shipment basis by container per RCP RW-4.

7. ATTACHMENTS

7.1 Form 69-10350, "Process Control Program (PCP) Verification," 06/03/93

7.2 "Major Change to the Solid Radwaste Treatment System Evaluation," 05/24/01

8. REFERENCES

8.1 Title 10 Code of Federal Regulations.

8.2 NUREG 0472 and 0473.

8.3 NUREG-0800, 11.4 US NRC Standard Review Plan Solid Waste Management Systems.

8.4 RCP.DC3, "Dewatering Control Program."

8.5 RP2.DC4, "Mobile Service Operating Procedure for Low-Level Radioactive WasteProcessing."

8.6 NRC Information Notice 88-08, Chemical Reactions with Radioactive WasteSolidification Agents.

8.7 ATG Nuclear Services, Quality Assurance Manual, Rev 1.

Document Control Master File, Catalog No. TK-9400/ATG-3.

8.8 Technical Position on Waste Form, Revision 1, US NRC, January 1991.

8.9 Cement Encapsulation of Cartridge Filters to Provide Waste Form Stability BasisDocument, Rev. 1, PG&E NRS Log 0087.

8.10 Encapsulation of Cartridge Filters In Vinyl Ester Styrene (VES) to Provide Waste FormStability Basis Document, Rev. 0, PG&E NRS Log 0072.

01180412 DOC OIB 1023 0812

|3 Ul/COIITROL ED PROCEDURE 2 DO NOT USERTO PERFORMW K/or ISSUEFOR USE a j;

69-10350 06/03/93 Page 1 of IDIABLO CANYON POWER PLANT

RP2.DC2ATTACHMENT 7.1

TITLE: Process Control Program (PCP) Verification

WASTE BATCH PCPDATE NOTES OPERATOR VERIFIERNumber Type Passed Failed

I7

I I II I

01180412 DOC 01B 1023 0812

h*c *f <UNCONTROLED PROCEDUE DO NOT USE TO PERFORM WORK or ISSUEFOR USEK** X

05/24/01 Page 1 of IDIABLO CANYON POWER PLANT

RP2.DC2ATTACHMENT 7.2

TITLE: Major Change to the Solid Radwaste Treatment System Evaluation

1. A summary of the evaluation that led to the determination that the change could be made in accordancewith 10 CFR 50.59;

2. Sufficient detailed information to totally support the reason for the change without benefit of additionalor supplemental information;

3. A detailed description of the equipment, components and processes involved and the interfaces withother plant systems;

4. An evaluation of the change which shows the predicted releases of radioactive materials in liquid andgaseous effluents and/or quantity of solid waste that differ from those previously predicted in theLicense application and amendments thereto;

5. An evaluation of the change which shows the expected maximum exposures to a MEMBER OF THEPUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previouslyestimated in the License application and amendments thereto;

6. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluent and insolid waste, to the actual releases for the period prior to when the changes are to be made;

7. An estimate of the exposure to plant operating personnel as a result of the change; and

8. Documentation of the fact that the change was reviewed and found acceptable.

01180412 DOC 011B 1023 0812


Attachment 5

Dewatering Control Program

(Procedure RP2.DC3, Revision 5)

V ** UN/CONITROLLED PROCEDUR D NTM USE TO PERFORM WORKor ISSUEFOR

PACIFIC GAS AND ELECTRIC COMPANY NUMBER RP2.DC3NUCLEAR POWER GENERATION REVISION 5DIABLO CANYON POWER PLANT PAGE 1 OF 4

ADMINISTRATIVE PROCEDURE

TITLE: Dewatering Control Program


PROCEDURE CLASSIFICATION: QUALITY RELATEDSPONSORING ORGANIZATION: RADIATION PROTECTION

REVIEW LEVEL: "A"

1. SCOPE

1.1 The purpose of the Radwaste Dewatering Control Program is to assure that all dewateringof vendor containers meets State and Federal Regulations as well as Burial Site Criteriafor Free Standing Water (FSW).

1.2 This Dewatering Process Control Program applies to all vendor dewatering liners andHigh Integrity Containers (HICs) containing ion exchange and filter media.

2. DISCUSSION

2.1 The dewatering system consists of a pump with necessary hoses for connection to theliner or HIC and to the plant interface piping. Liners or HICs supplied by a vendor arepreassembled with a hub and/or lateral assembly for dewatering and media retention.Procedures for dewatering are dependent on the waste material to be dewatered and/or thecontainer type.

2.2 These procedures, specific to the container and waste material provide instruction for thepumping and settling time sequences as well as; pump rate, total times pumped, andacceptance criteria. Additional contingencies are provided for vessels or liners requiringfurther steps to meet acceptance criteria and the means of verification of compliance.Methods for recording dewatering data are included in the specific procedure.

3. RESPONSIBILITIES

3.1 The radiation protection manager is responsible for implementation of the requirements ofthis program.

3.2 The radwaste engineer is responsible for the development and review of proceduresrelating to the requirements of this program.

3.3 The radwaste foreman is responsible for implementation of the procedures relating to therequirements of this program.

3.4 Quality organization is responsible for verification of compliance with the Qualityrequirements of the dewatering procedures.

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5.

*L -: UA t CONTRO£LED PROCEflURF-D9NOTUSE O PERFORM WORK r& ISSUEFOR USE BOGPACIFIC GAS AND ELECTRIC COMPANY NUMBER RP2.DC3DIABLO CANYON POWER PLANT REVISION 5

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4. PREREQUISITES

4.1 In order to stabilize a dewatered wet waste an NRC approved HIC can be utilized.

4.2 Polyethylene HICs which are listed in the U.S. Ecology Ward Valley license may be usedto provide waste stability provided they are disposed of in a concrete overpack as acompensatory measure for their structural deficiency.

5. PRECAUTIONS

5.1 Any container of dewatered ion exchange media placed in interim storage shall undergo adewatering verification prior to shipment for disposal.

5.2 No NRC Topical or vendor dewatering test report for spent cartridge filters currentlyexists. Spent cartridge filters may be loaded into HICs and dewatered for storage. Thesecontainers shall not be shipped for disposal prior to encapsulation of the contents ordewatering by an as yet to be determined verifiable process to meet Free Standing Watercriteria.

6. INSTRUCTIONS

6.1 The Dewatering Control Program for Liners and HICs, shall consist of proceduresreviewed and approved by the Station Director.

6.2 These procedures shall contain the specific instructions for:

6.2.1 Pumping Time Cycles

6.2.2 Pump Rate

6.2.3 Total Time to be Pumped

6.2.4 Temperature monitoring

6.2.5 Acceptance Criteria of FSW

6.2.6 Pressure Monitoring when Gas Generation is Suspected

6.2.7 Documentation and Records of Dewatering Activities

6.3 These procedures shall also specify the type of container and waste media that may bedewatered. Contingencies are provided for additional steps that may be necessary to meetFSW criteria.

6.4 GENERAL CONTROL PARAMETERS

6.4.1 Carbon Steel liners for direct disposal at Barnwell are not to be loaded withmaterial over I pCi/cc of nuclides with half lives greater than five years.Liners for direct disposal at Envirocare are not to be loaded with materialgreater than Class A.

6.4.2 Waste temperature is monitored to mitigate exothermic chemical reactionswhich are possible when dewatering organic ion exchange materials.

6.4.3 At the end of the appropriate pump cycle, as outlined by the procedure, ameasurement of displaced liquid is made.

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TITLE: Dewatering Controi Program

6.4.4 Records as outlined by the dewatering procedure shall be maintained for alldewatered containers prior to shipment. Copies are delivered to the RadwasteForeman for indexing with radwaste container paper work.

6.5 Reports

The following are maintained in Document Control Master File, CatalogNo. TK 9400/SEG-1.

6.5.1 Test Report: LN Technologies, Corp. LN 87-008, DB Series DewateringContainers, Rev. A.

6.5.2 Test Report: LN Technologies, Corp. EN-009, Powdered Resin Dewateringwith a Sandpiper Pump, Rev. A.

6.5.3 Test Report: Scientific Ecology Group, Inc. EN-005, Summary Test Reportfor Dewatering Tests on Durasil 70, DT-80 and other media DewateringTesting, Rev. C.

6.5.4 Topical Report for the SEG Composite High Integrity Container LN-87-005,Rev. 1, March 1990, Docket Number WM-93.

6.5.5 NRC Technical Evaluation Report LN Composite High-Integrity ContainerManufactured by LN Technologies Corporation, Docket Number WM-93.

6.5.6 Waste characterization: MMT of Tennessee, PGE-1, IXE800G.

6.5.7 State of South Carolina Department of Health and Environmental Control,Certificate of Compliance- High Integrity Container.

a. Certification Number DHEC-HIC-ML0 16

b. Certification Number DHEC-PL-014

c. Certification Number DHEC-PL-0 10

d. Certification Number DHEC-PL-007

7. RECORDS

7.1 Records of dewatering activities shall be submitted to the Records Management Systemby container.

8. ATTACHMENTS

None

9. REFERENCES

9.1 RP2.DC4, "Mobile Service Operating Procedure for Low-Level Radioactive WasteProcessing."

9.2 NRC Information Notice No. 83-14: Dewatered Spent Ion Exchange Resin Susceptibilityto Exothermic Chemical Reaction.

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9.3 NRC Information Notice No. 89-27: Limitations on the Use of Waste Forms and HighIntegrity Containers for the Disposal of Low-Level Radioactive Waste.

9.4 NRC Information Notice No. 90-50: Minimization of Methane Gas in Plant Systems andRadwaste Shipping Containers.

9.5 Polyethylene HIC and Concrete Overpack Disposal Basis Document, Rev 1, PG&E NRSLog 0041.

9.6 Radioactive Material License issued by the State of California to U.S. Ecology for WardValley, 4505-36, Sept. 16, 1993.

9.7 ATG Nuclear Services, Quality Assurance Manual, Rev 1, Document Control MasterFile, Catalog No. TK-9400/ATG-3.

10. SPONSOR

C. Clint Miller

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Attachment 6

2002 Land Use Census


2002 LAND USE CENSUS

Diablo Canyon Pbwer Plant (DCPP) radiation protection personnel conducted aland use census in the vicinity of DCPP for 2002. The land use census is based onNuclear Regulatory Commission (NRC), Regulatory Guide 4.8, "EnvironmentalTechnical Specifications for Nuclear Power Plants", and required by DCPP ProgramDirective CY2, 'Radiological Monitoring and Controls Program." The census isconducted at least once per year during the growing season for the Diablo Canyonenvirons.

The land use census identifies the nearest milk animal and broadleaf producinggarden greater than 50 square meters (500 square feet) in each of the landwardmeteorological sectors within a distance of 8 kilometers (5 miles) of the plant.- DCPPIDAP RP1.ID11, 'Environmental Radiological Monitoring Program", requires that thenearest residence be identified in each of the landward sectors within a distance of 5miles.

The land use census was performed by directly contacting individual landownersI tenants and by aerial surveys. The landowners or tenants were contacted betweenSeptember 26th and October 9th, 2002. The aerial survey was performed on September30th, 2002.

,, The census identified one household garden greater than 50 square meters(500 square feet) that produces broadleaf vegetation in the East sector at 4.5 miles fromDCPP Unit 1. No milk animals were identified within the first 5 miles of any sector.Much of the area surrounding the plant site is used for rotational cattle grazing by twoseparate cattle companies. Various numbers of cattle are sold to market at the end ofeach year. One cattle operation utilizes the land north of the plant site. A second cattleoperation utilizes the land south of the plant site. The rancher for the northern cattleoperation slaughters about 2 cattle per year for personal consumption. Goats were usedfor weed abatement for approximately 6 months within the area surrounding the plantsite. The rancher from the northern operation owns these goats and also slaughtersabout 2 goats per year for personal consumption. During 2002, approximately 100 goatsare to be sold in mass-market auction. A farm is located on the coastal plateau, alongthe site access road, in the east-southeast (ESE) sector. The farm starts atapproximately 3.3 miles and extends to 4.5 miles from the plant. This commercial farmproduces 75% legumes (sugar peas) and 25% cereal grass (oat hay).

A total of 13 residences were identified within the 5-mile radius of the plant that wereconfirmed or appear to be occupied during 2002. Two abandoned structures are locatedat 1.6 miles north-northwest (NNW) of the plant. The nearest residence, relative to allsectors, is a small trailer 1.2 miles northwest (NW) of the plant (occupied approximately1 month per year). Ranchers use this trailer during cattle round-ups. Table 1summarizes the results of the land use census and Figure 1 shows the locations of thefarm, garden, and residences in the vicinity of DCPP.


Table I

Land Use Census 2002

Distance in Miles from the Unit 1 Center Line to theNearest Milk Animal, Residence, Vegetable Garden

221/2 Degree(a) Nearest Nearest Residence NearestRadial Sector Milk Animal Residence Azimuth Vegetable

km (mi) Degree Gardenkm (mi)

NW None 1.93 (1.2) 319.5 None

NNW None 2.41 (1.5) 331 None

N None None None

NNE None 5.3 (3.3) 018.5 None

NE None 7.89 (4.9) 036 None

ENE None 7.08 (4.4) 063.5 None

E None 5.95 (3.7) 097.5 None

E None 7.24 (4.5) 098 7.24 (4.5)

ESE None None 5.28 (3 .3 )(b)

SE None None None

Table Notation:

(a) Sectors not shown contain no land beyond the site boundary, otherthan islets not used for the purposes indicated in this table.

(b) The vegetable garden indicated is the commercial farm along the westwardside of the site access road; however, it does not produce broadleafvegetation. Area is about 100 acres of land with 6 to 10 rotational plantingsper year (not all 100 acres planted at any one time). Commercial crops areabout 75% sugar peas and 25% oat hay. The farm starts at approximately 3.3miles and extends to 4.5 miles from the plant.


NW

UNITS 1 AND 2DIABLO CANYON SITE

El Gardens or Farm

A Residences

970338ilanduse 02p

N,

0 1 2 3 4 5SCALE IN MILES

Figure 1. Units I and 2 Diablo Canyon Povuer Plant Land Use Census.